What type of reactor is Watts Bar Unit 2?

Watts Bar Nuclear Power Plant Units 1 & 2 cooling towers and containment buildings. The plant, construction of which began in 1973, has two Westinghouse pressurized water reactor units: Unit 1, completed in 1996, and Unit 2, completed in 2015.

How much power can a small modular reactor produce?

Microreactors are 100 to 1,000 times smaller than conventional nuclear reactors, while small modular reactors (SMRs) range from 20 to 300 megawatts . Microreactors offer a combination of reliability and operational flexibility that no other small generating system can match.

Is RBMK a PWR or BWR?

The RBMK-1000 reactor The RBMK-1000 (Figure 2) is a Soviet designed and built graphite moderated pressure tube type reactor , using slightly enriched (2% 235U) uranium dioxide fuel. It is a boiling light water reactor, with direct steam feed to the turbines, without an intervening heat-exchanger.

What type of reactor is Hinkley Point?

Hinkley Point C nuclear power station is a project to construct a new 3,200 MWe nuclear power station and is the first to be built in a generation (since Sizewell B in 1995). The two EPR1 reactors aim to provide the UK with low-carbon electricity for around six million homes over its 60 year expected life span.

What type of reactor is Hinkley Point B?

Hinkley Point B nuclear power station Reactor type GCR - AGR Reactor supplier General Electric Company, Whessoe, and Strachan & Henshaw (Graphite supplied by Anglo Great Lakes Corporation Ltd) Cooling source Bridgwater Bay Thermal capacity 2 x 1494 MW t 25 more rows

What type of reactor is AP1000?

The AP1000® reactor is a two-loop pressurized water reactor (PWR) that uses a simplified, innovative, and effective approach to safety.

CNL-20-006, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01) (2024)

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EPID:L-2020-LLA-0223, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01) (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Withholding Request Acceptance Questions 1 April 2021 Answers 29 April 2021 Results Approval Other:
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Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01)
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ML20276A092
Site:	Watts Bar
Issue date:	10/02/2020
From:	Jim Barstow Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML20276A091	List: CNL-20-006, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01) ML21021A241
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Category:Letter type:CNL [Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50 [Table view]

{{#Wiki_filter:Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-006 October 2, 2020 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating Licenses Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391

Subject:

Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 "Containment Leakage Rate Testing Program" (WBN-TS-19-01)

Reference:

1. Nuclear Energy Institute 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"

dated July 2012, (ML12221A202)

2. Electrical Power Research Institute Report, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325," dated October 2008

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk- Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018 (ML17317A256)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License Nos. NPF-90 and NPF-96 for the Watts Bar Nuclear Plant (WBN),Units 1 and 2, respectively.The proposed changes would revise the WBN Units 1 and 2, Technical Specifications (TS) 5.7.2.19, "Containment Leakage Rate Testing Program," by adopting Nuclear Energy Institute (NEI) 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," (Reference 1) as the implementation document for the performance-based Option B of 10 CFR Part 50, Appendix J. The proposed changes permanently extend the Type A containment integrated leak rate testing (CILRT) interval from 10 years to 15 years and the Type C local leakage rate testing intervals from 60 months to 75 months. In addition, a clarification of the value of Pa to be used for containment leakage rate testing purposes is incorporated.Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3 U.S. Nuclear Regulatory Commission CNL-20-006 Page 2 October 2, 2020 The proposed amendment is considered risk-informed. An evaluation has been performed to assess the risk effect of the proposed change. The risk assessment follows the guidelines of Reference 1, and the corresponding Electrical Power Research Institute (EPRI) Report, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" (Reference 2), and the guidance of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 3). to this submittal provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachments 1 and 2 to the enclosure provide the existing TS pages marked-up to show the proposed changes for WBN Unit 1 and Unit 2, respectively.Attachments 3 and 4 to the enclosure provide the WBN Unit 1 and Unit 2 TS pages retyped to show the proposed changes. Attachment 5 to the enclosure provides the existing WBN Unit 1 TS Bases pages marked-up to show the proposed changes. Only the Unit 1 TS Bases pages have been provided, as the Unit 2 changes will be nearly identical except for some editorial differences. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program. to this submittal is a probabilistic risk assessment (PRA) evaluation for permanently extending the containment Type A test interval. Enclosure 2 provides resolutions of the PRA Peer Review Team Facts & Observations and impact on the proposed license amendment. to this submittal is the Kalsi Engineering Report 3960C, Evaluation of Higher Test Pressure on Leakage for Watts Bar, in support of the value of Pa to be used for containment leakage rate testing purposes. Enclosure 3 contains information that Kalsi Engineering considers to be proprietary in nature in accordance with 10 CFR Section 2.390. contains a non-proprietary version of Enclosure 3. Enclosure 5 provides the affidavit from Kalsi Engineering supporting the proprietary withholding request. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission (NRC) and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390. Accordingly, TVA requests that the information which is proprietary to Kalsi Engineering be withheld from public disclosure in accordance with 10 CFR Section 2.390. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Kalsi Engineering affidavit should be addressed to Neal Estep, Senior Vice President, Kalsi Engineering, Inc., 745 Park Two Drive, Sugar Land, TX 77478.TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the TS changes qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Environment and Conservation.Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3 U.S. Nuclear Regulatory Commission CNL-20-006 Page 3 October 2, 2020 TVA requests approval of the proposed license amendment within one year from the date of this submittal with implementation within 30 days following NRC approval.There are no new regulatory commitments associated with this submittal. If you have any questions about this proposed change, please contact Gordon Williams, Senior Manager, Fleet Licensing (Acting) at (423) 751-2687.I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2nd day of October 2020.Respectfully, James Barstow Vice President, Nuclear Regulatory Affairs & Support Services Enclosures

1. Evaluation of Proposed Change

2. PRA Evaluation

3. Kalsi Engineering Report 3960C (Proprietary)

4. Kalsi Engineering Report 3960C (Non-Proprietary)

5. Kalsi Engineering Affidavit cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 3

Enclosure 1 Evaluation of Proposed Change

Subject:

Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 "Containment Leakage Rate Testing Program" (WBN-TS-19-01)Table of Contents 1.0

SUMMARY

DESCRIPTION........................................................................................ 3 2.0 DETAILED DESCRIPTION ........................................................................................ 3

3.0 TECHNICAL EVALUATION

....................................................................................... 4 3.1 Background 4 3.2 Description of WBN Containment 6 3.3 Leak Test History 7 3.3.1 Type A Integrated Leak Rate Test .................................................................... 7 3.3.2 Type B and Type C Testing............................................................................... 9 3.4 Containment Inspections 13 3.4.1 Containment Inservice Inspections ................................................................. 13 3.4.2 Containment Coatings Inspections .................................................................. 21 3.5 Industry Operating Experience Review 22 3.6 NRC Limitations and Conditions for NEI 94-01 23 3.6.1 June 25, 2008 NRC Safety Evaluation ............................................................ 23 3.6.2 June 8, 2012 NRC Safety Evaluation .............................................................. 25 3.7 Plant-Specific Confirmatory Analysis 28 3.7.1 Methodology.................................................................................................... 28 3.7.2 Probabilistic Risk Assessment (PRA) Acceptability ......................................... 30 3.7.3 Conclusions of the Plant-Specific Risk Assessment Results ........................... 31 3.8 Basis for the Proposed TS Changes 32 3.8.1 General basis .................................................................................................. 32 3.8.2 Use of a bounding value for Pa ........................................................................ 33 3.8.3 Deviation #1 from ANSI/ANS 56.8-2002 related to the use of a bounding Pa.. 33 3.8.4 Deviation #2 from ANSI/ANS 56.8-2002 related to the use of a bounding Pa.. 34 3.9 Conclusion 35 CNL-20-006 E1-1 of 41

Enclosure 1

4.0 REGULATORY EVALUATION

................................................................................. 36 4.1 Applicable Regulatory Requirements and Criteria 36 4.1.1 Regulations ..................................................................................................... 36 4.1.2 General Design Criteria ................................................................................... 36 4.2 Precedent 37 4.3 No Significant Hazards Consideration 37 4.4 Conclusions 39

5.0 ENVIRONMENTAL CONSIDERATION

.................................................................... 40

6.0 REFERENCES

......................................................................................................... 40 Attachments

1. Proposed TS Changes (Mark-Ups) for WBN Unit 1

2. Proposed TS Changes (Mark-Ups) for WBN Unit 2

3. Proposed TS Changes (Final Typed) for WBN Unit 1

4. Proposed TS Changes (Final Typed) for WBN Unit 2

5. Proposed TS Bases Page Changes (Mark-Ups) for WBN Unit 1 (For Information Only)

CNL-20-006 E1-2 of 41

Enclosure 1 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) is requesting a license amendment to Facility Operating License Nos. NPF-90 and NPF-96 for the Watts Bar Nuclear Plant (WBN), Units 1 and 2.This evaluation supports a request to revise WBN Units 1 and 2, Technical Specifications (TS) 5.7.2.19, "Containment Leakage Rate Testing Program," by adopting Nuclear Energy Institute (NEI) 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"(Reference 1) as the implementation document for the performance-based Option B of 10 CFR Part 50, Appendix J. The proposed changes permanently extend the Type A containment integrated leak rate testing (CILRT) interval from 10 years to 15 years and the Type C local leakage rate testing (LLRT) intervals from 60 months to 75 months. In addition, a clarification of the value of Pa to be used for containment leakage rate testing purposes is incorporated.2.0 DETAILED DESCRIPTION The following is a detailed description of the proposed TS changes for WBN Unit 1 and Unit 2.WBN Unit 1 TS 5.7.2.19 is revised as follows:5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995 NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 15.0 psig.For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound a range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.CNL-20-006 E1-3 of 41

Enclosure 1 WBN Unit 2 TS 5.7.2.19 is revised as follows:5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995.NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound thea range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.Note that the existing version of TS 5.7.2.19 for WBN Unit 2 already included the language regarding the use of 15.0 pounds per square inch gauge (psig) as Pa for containment leakage rate testing program purposes.Attachments 1 and 2 to this enclosure provide the existing TS pages marked-up to show the proposed changes for WBN Unit 1 and Unit 2, respectively. Attachments 3 and 4 to this enclosure provide the WBN Unit 1 and Unit 2 TS pages retyped to show the proposed changes. Attachment 5 to this enclosure provides the existing WBN Unit 1 TS Bases pages marked-up to show the proposed changes. Only the Unit 1 TS Bases pages have been provided, as the Unit 2 changes will be nearly identical except for some editorial differences. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.

3.0 TECHNICAL EVALUATION

3.1 Background

The testing requirements of 10 CFR 50, Appendix J (Reference 2), provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in TS 5.7.2.19. The periodic surveillance of containment penetrations and isolation valves ensure that proper maintenance and repairs can be performed on the systems and components penetrating containment during the service life of the containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests: (1) Type A tests, intended to measure the containment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for containment penetrations; and (3) Type C tests, intended to measure containment isolation valve leakage. Type B and C tests identify the vast majority of potential containment leakage paths. Type A tests identify the overall CNL-20-006 E1-4 of 41

Enclosure 1 (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Type B and Type C testing.This request modifies the existing Appendix J Type A and Type C testing intervals but does not change the Appendix J Type A or Type C test methods. The CILRT imposes significant expense on the station while the safety benefit of performing this test within 10 years, versus 15 years, is minimal. Increasing the allowable extended testing interval for Type C LLRTs by 15 months will result in a reduction in the amount of testing required, with commensurate reductions in radiation exposure, personnel time in lining up for tests, draining systems, conducting tests, and the risk involved in performing such testing while the safety benefit of performing Type C LLRTs within 60 months, versus 75 months, is minimal. This request represents a cost-beneficial licensing change with minimal effect on safety margin.In 1995, 10 CFR 50, Appendix J was amended to provide a performance-based Option B for containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals, as well as to the criteria necessary to meet the requirements of Option B. Also in 1995, NRC Regulatory Guide (RG) 1.163 (Reference 3) was issued. RG 1.163 endorsed NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 94-01, Appendix J" (Reference 4), with certain modifications and additions.Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory integrated leak rate test (ILRT) performance history (i.e., two consecutive successful Type A tests) to reduce the test frequency from the containment Type A ILRT test from three tests in ten years to one test in ten years.This relaxation was based on an NRC risk program documented in NUREG-1493, "Performance-Based Containment Leak-Test Program" (Reference 5) and Electric Power Research Institute (EPRI) Topical Report (TR)-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals" (Reference 6),both of which illustrated that the risk increase associated with extending the ILRT surveillance interval was very small.NEI 94-01, Revision 2 (Reference 7), describes an approach for implementing the optional performance-based requirements of Option B described in 10 CFR 50, Appendix J, which includes provisions for extending Type A intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. NEI 94-01, Revision 2 delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance data, plant-specific performance data, and risk insights in determining the appropriate testing frequency. NEI 94-01, Revision 2, also discusses the performance factors that licensees must consider in determining test intervals. However, NEI 94-01, Revision 2 does not address how to perform the tests because these details are included in existing documents (e.g., American National Standards Institute/American Nuclear Society [ANSI/ANS]-56.8-2002). The NRC final Safety Evaluation (SE), issued by letter dated June 25, 2008 (Reference 8), documents the NRC's evaluation and acceptance of NEI 94-01, CNL-20-006 E1-5 of 41

Enclosure 1 Revision 2, subject to the specific limitations and conditions listed in Section 4.1 of the SE. The accepted version of NEI 94-01 was subsequently issued as Revision 2-A dated October 2008 (Reference 9).EPRI Report 1009325, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," Revision 2 (Reference 10), provides a risk impact assessment for optimized ILRT intervals of up to 15 years, utilizing current industry performance data and risk-informed guidance. The assessment validates increasing allowable extended LLRT intervals to the 120 months as specified in NEI 94-01, Revision 0.However, the industry requested that the allowable extended interval for Type C LLRTs be increased only to 75 months, to be conservative, with a permissible extension (for non-routine emergent conditions) of nine months (i.e., 84 months total). The NRC's final SE, issued by letter dated June 25, 2008, documents the NRC's evaluation and acceptance of EPRI Report 1009325, Revision 2, and extension of the Type C LLRT interval to 75 months, subject to the specific limitations and conditions listed in Section 4.2 of the SE. An accepted version of EPRI Report 1009325, Revision 2 was subsequently issued as EPRI Report 1018243, "Risk Impact of Extended Integrated Leak Rate Testing Intervals - Revision 2-A of 1009325," dated October 2008 (Reference 12).NEI 94-01, Revision 3 (Reference 13), describes an approach for implementing the optional performance-based requirements of Option B described in 10 CFR 50, Appendix J, which includes provisions for extending Type A and Type C intervals to up to 15 years and 75 months, respectively, and incorporates the regulatory positions stated in RG 1.163. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. In determining the appropriate testing frequency, this method uses industry performance data, plant-specific performance data, and risk insights, primarily Revision 3 of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Bases" (Reference 11). NEI 94-01, Revision 3, also discusses the performance factors that licensees must consider in determining test intervals. However, it does not address how to perform the tests because these details are included in existing documents (e.g., ANSI/ANS-56.8-2002). The NRC final SE issued by letter dated June 8, 2012 (Reference 14), documents the NRC's acceptance of NEI 94-01, Revision 3, subject to the specific limitations and conditions listed in Section 4.0 of the SE. The accepted version of NEI 94-01 was subsequently issued as Revision 3-A dated July 2012 (Reference 1).3.2 Description of WBN Containment The WBN containment consists of a containment vessel and a separate Shield Building enclosing an annulus. The containment vessel is a freestanding, welded steel structure with a vertical cylinder, hemispherical dome, and a flat circular base.The Shield Building is a reinforced concrete structure similar in shape to the containment vessel.The WBN steel containment vessel (SCV) is a low-leakage, freestanding steel structure consisting of a cylindrical wall, a hemispherical dome, and a bottom liner plate encased in concrete. The structure consists of side walls measuring 114 feet 8-5/8 inches in height from the liner on the base to the spring line of the CNL-20-006 E1-6 of 41

Enclosure 1 dome and has an inside diameter of 115 feet. The bottom liner plate is 1/4 inch thick, the cylinder varies from 1-3/8 inch thickness at the bottom to 1-1/2 inch thick at the springline, and the dome varies between 1-3/8 inch thickness and 13/16 inch thickness with 15/16 inch thickness at the apex.The SCV is provided with both circumferential and vertical stiffeners on the exterior of the shell. These stiffeners are required to satisfy design requirements for expansion and contraction, seismic forces, and pressure transient loads. The circumferential stiffeners were installed on approximately 10-foot centers during erection to ensure stability and alignment of the shell. Vertical stiffeners are spaced at 5° between the two lowest circumferential stiffeners. Other locally stiffened areas are provided at the equipment hatch and two personnel locks.During the WBN Unit 1 steam generator replacement in Fall 2006, two construction openings were made in the steel containment vessel. These openings were restored by reinstalling the removed steel sections and rewelding them to the remaining structure using full penetration welds. Abandoned-in-place reinforcement and support members to stiffen the SCV during creation and use of the two construction openings are designed to remain attached to the SCV during a seismic event. The integrity of the restored vessel was verified by nondestructive examination (NDE) and leak testing of the welds.The WBN dual-unit Updated Final Safety Analysis Report (UFSAR) Section 3.8.2.2.2 describes that the SCV maximum internal pressure is 15 psig at 250°F and the design internal pressure is 13.5 psig at 250°F. The difference between the two is explained in the same section by stating, "Paragraph NE-3312(b) of Section III of the ASME Code states that the 'design internal pressure' of the vessel may differ from the 'maximum containment pressure', but in no case shall the design internal pressure be less than 90% of the maximum containment internal pressure."3.3 Leak Test History 3.3.1 Type A Integrated Leak Rate Test WBN TS 5.7.2.19 requires the measurement of the containment leakage rate.TS 5.7.2.19 limits as-left Type A leakage to 0.75 La. The results of past Type A tests for WBN are provided in Tables 3.3.1-1 and 3.3.1-2. The method for leakage determination is the mass point 95 percent upper confidence level (UCL) estimate of leakage rate. Previous ILRT testing confirmed that WBN Units 1 and 2, containment structures leakage is acceptable, with considerable margin, with respect to the TS acceptance criteria of 0.25% of containment air weight per day at the design basis loss of coolant accident (LOCA) pressure (Pa). Because the WBN Unit 1 and WBN Unit 2 Type-A results (as shown below) meet the performance leakage rate criteria from NEI 94-01, Revision 3-A, a test frequency of at least once per 15 years would be in accordance with NEI 94-01, Revision 3-A.CNL-20-006 E1-7 of 41

Enclosure 1 Table 3.3.1-1 WBN Unit 1 As Found Leak Rate Acceptance Limits Test Pa Pt Pd Date psig psig psig Method Result As- As-Left Found 10/12/12 15 15.03 13.5 Mass Point UCL 0.0904135%/day 0.25%/day 0.1875%/day leakage with (1.0 La) (0.75 La) penalties 10/10/97 15 14.97 13.5 Mass Point UCL 0.10613%/day 0.25%/day 0.1875%/day leakage with (1.0 La) (0.75 La) penalties U1 15 14.59 13.5 Mass Point UCL 0.01683%/day 0.25%/day 0.1875%/day Startup* leakage with (1.0 La) (0.75 La)(06/23/94) penalties Pa - As defined in WBN U1 TS 5.7.2.19 Pt - Final Test Pressure (psig) - minimum allowable Pt is Pa -1 psig = 14 psig Pd - Containment Design Pressure La - 0.25% containment air weight per day at Pa

 * - value from startup is as-left data Table 3.3.1-2 WBN Unit 2 As Found Leak Rate Acceptance Limits Pa Pt Pd Test Date psig psig psig Method Result As- As-Left Found 05/09/19 15 15.26 13.5 Mass Point UCL 0.0590%/day 0.25%/day 0.1875%/day leakage with (1.0 La) (0.75 La) penalties U2 15 15.32 13.5 Mass Point UCL 0.01399%/day 0.25%/day 0.1875%/day Startup* leakage with (1.0 La) (0.75 La)

(08/27/15) penalties Pa - As defined in WBN U1 TS 5.7.2.19 Pt - Final Test Pressure (psig) - minimum allowable Pt is Pa -1 psig = 14 psig Pd - Containment Design Pressure La - 0.25% containment air weight per day at Pa

 * - value from startup is as-left data In accordance with NEI 94-01, Revision 3-A (Reference 1), Section 9.1.2, further extensions in test intervals are based upon two consecutive, periodic, successful Type A tests and requirements stated in Section 9.2.3 of this guideline. The results in Tables 3.3.1-1 and 3.3.1-2 show that there has been margin to the maximum allowable leakage rate of 0.25 wt%/day in the last two consecutive successful Type A tests for each unit.

There is no anticipated addition or removal of plant hardware within containment that could affect leak-tightness that would not be challenged by local leak rate CNL-20-006 E1-8 of 41

Enclosure 1 testing. There are no known modifications that will require a Type A test to be performed prior to Fall 2027 for Unit 1 and Spring 2034 for Unit 2, when the next Type A tests are due for performance in accordance with this proposed change.TVA plans to replace the WBN Unit 2 Steam Generators during the Unit 2 Cycle refueling outage (U2R5 scheduled for Fall 2023). A local leak inspection of the applicable containment dome welds using bubble solution while the primary containment is pressurized to at least Pa will be performed during U2R5 as post-maintenance testing (PMT) in accordance with the condition on use of NEI 94-01, Revision 2-A. Any unplanned modifications to the containment prior to the next scheduled Type A test would be subject to the testing requirements of NEI 94-01, Revision 3-A, Section 9.2.4, as applicable, and as discussed in condition 4 from the June 25, 2008 safety evaluation (Reference 8) for NEI 94-01, Revision 2 (see Table 3.6.1-1).3.3.2 Type B and Type C Testing As discussed in NUREG-1493, Type B and Type C tests can identify the vast majority of all containment leakage paths. This amendment request adopts the guidance in NEI 94-01, Revision 3-A in place of NEI 94-01, Revision 0, but otherwise does not affect the scope, performance or scheduling of Type B or Type C tests.Type B and Type C testing will continue to provide a high degree of assurance that containment leakage rates are maintained well within limits.The WBN Units 1 and 2, Appendix J, Type B and Type C leakage rate test program requires testing of electrical penetrations, airlocks, hatches, bellows, and valves within the scope of the program as required by 10 CFR 50, Appendix J, Option B and TS 5.7.2.19. The Type B LLRTs for each unit includes; two personnel airlocks, 46 individual bellows, 53 electrical penetrations, 12 resilient seals, and an Equipment Hatch. The Type C LLRTs for Unit 1 include 81 valve penetrations and the Type C LLRTs for Unit 2 include 73 valve penetrations.CNL-20-006 E1-9 of 41

Enclosure 1 Summary of Recent Type B and C Testing A review of the Type B and Type C test results from the recent WBN Unit 1 and Unit 2 refueling outages in Tables 3.3.2-1 and 3.3.2-2 demonstrates a large margin between the actual as-found and as-left outage summations and the TS leakage rate acceptance criteria (that is, less than 0.6 La).Table 3.3.2-1 Unit 1 Type B and Type C LLRT Totals As-Found Minimum As-Left Maximum Pathway Pathway Refueling Percentage of Percentage of Outage Total Total 0.6La 0.6La Leak Rate Leak Rate (147.6 scfh) (147.6 scfh)U1R16 8.97 scfh 6.1% 22.44 scfh 15.2%05/31/20 U1R15 6.78 scfh 4.6% 20.84 scfh 14.1%10/18/18 U1R14 5.30 scfh 3.6% 15.31 scfh 10.4%04/23/17 U1R13 5.57 scfh 3.8% 16.60 scfh 11.2%10/16/15 Table 3.3.2-2 Unit 2 Type B and Type C LLRT Totals As-Found Minimum As-Left Maximum Pathway Pathway Refueling Percentage of Percentage of Outage Total Total 0.6La 0.6La Leak Rate Leak Rate (147.6 scfh) (147.6 scfh)U2R2 19.75 scfh 13.4% 31.87 scfh 21.6%05/10/19 U2R1 12.50 scfh 8.5% 31.02 scfh 21.0%11/27/17 U2 Startup n/a n/a 15.86 scfh 10.7%03/15/16 CNL-20-006 E1-10 of 41

Enclosure 1 Repeat LLRT Failures over the Last Two Refueling Outages A repeat failure is defined as two consecutive failures of the as-found LLRT administrative limit. There have been three repeat failures in Unit 1 and three repeat failures in Unit 2 over the last two outages. In each case, the failure was a Type C tested valve as described in Tables 3.3.2-3 and 3.3.2-4.Table 3.3.2-3 Unit 1 Repeat LLRT Failures Valve Outage As-Found Admin As-Left Discussion

 / Date Leak Rate Limit Leak Rate Residual foreign material was observed in the seat area. Valve was cleaned U1R16 1.25 scfh 1.0 scfh 0.0 scfh and reworked. New disc and stem internals were 1-FCV-90-116 installed Foreign material was observed in the seat area.

U1R15 3.259 scfh 1.0 scfh 0.034 scfh Seat area was cleaned and new diaphragm and packing installed No maintenance performed U1R16 1.49 scfh 1.0 scfh 1.49 scfh based on no active adverse trend present.1-FCV-63-23 No maintenance performed U1R15 1.38 scfh 1.0 scfh 1.38 scfh based on no active adverse trend present.No maintenance performed based on Post Accident Sampling (PAS) valves U1R16 3.41scfh 1.0 scfh 3.41 scfh maintained in the block position with no possibility 1-FCV-43-287 for adverse trend possible.No maintenance performed based on PAS valves U1R15 5.28 scfh 1.0 scfh 5.28 scfh maintained in the block position with no possibility for adverse trend possible.CNL-20-006 E1-11 of 41

Enclosure 1 Table 3.3.2-4 Unit 2 Repeat LLRT Failures Valve Outage As-Found Admin As-Left Discussion

 / Date Leak Rate Limit Leak Rate Residual foreign material was observed in the seat area. Valve was cleaned U2R2 64.50 scfh 1.0 scfh 0.39 scfh and reworked. New disc and stem internals were 2-CKV-31-3392 installed.

Foreign material was observed in the seat U2R1 Gross 1.0 scfh 0.005 scfh area. Valve was disassembled cleaned and inspected.No maintenance performed based on no U2R2 4.60 scfh 2.0 scfh 4.60 scfh active adverse trend present from U2R1 as-left 2-FCV-32-111 test to U2R2 as-found test.Attempts at fixing the valve by machining the U2R1 2.20 scfh 2.0 scfh 4.39 scfh plug seat were performed with limited success Discovered loose ceramic isolator.U2R2 7.968 scfh 0.5 scfh 0.07 scfh Re-torqued isolator to Electrical correct leakage.penetration X- No maintenance 122E performed. Performed U2R1 4.43 scfh 0.5 scfh 4.43 scfh leakage evaluation to allow leakage to remain as-is.Extension of Type B and C Testing Extensions of the Type B and Type C testing intervals are governed by NEI 94-01 and the associated regulatory documents that endorse its use. WBN Units 1 and 2 currently comply with NEI 94-01, Revision 0 in accordance with RG 1.163 which allow the Type B and Type C testing intervals for most components to be extended out to 120 months and 60 months, respectively, based on acceptable as-found LLRT performance and other factors. Table 3.3.2-5 provides a breakdown of those components currently on NEI 94-01, Revision 0 or RG 1.163 extended frequency by unit and test type.Table 3.3.2-5 Unit 1 Totals Unit 2 Totals LLRT Extended LLRT Extended Components Frequency Components Frequency Type B 114 114 114 113 Type C 189 161 184 146 CNL-20-006 E1-12 of 41

Enclosure 1 Following each refueling outage (18-month cycle), LLRT results are reviewed, component performance is evaluated, and LLRT frequencies are changed accordingly - reduced for failure to meet administrative limits and extended, if eligible, following two successful as-found tests. Results of these evaluations and frequency changes are summarized in a post-outage report and this information is used for development of future outage scope of periodic as-found LLRTs.Development of outage scope for conditional as-found and as-left LLRTs is based on planned maintenance and modification activities that have potential to affect containment leakage integrity.There is adequate margin in the percentage of total allowable leak rate to accommodate addition of an understatement for those Type C tested valves extended from 60 months to 75 months as required by the condition on use of NEI 94-01, Revision 3-A.The number of LLRTs on extended frequency and the low percentage of the total of allowable leak rate used demonstrate the WBN LLRT program will provide continuing assurance that the most likely sources of leakage will be identified and repaired.3.4 Containment Inspections 3.4.1 Containment Inservice Inspections The WBN Containment Inservice Inspection (CISI) Program provides the site-specific requirements for implementation of the examinations of the SCV, American Society of Mechanical Engineers (ASME) Code Class MC components, in accordance with ASME Code, Section XI, Subsection IWE, 2013 Edition, as conditioned by 10 CFR 50.55a.The WBN Unit 2 First Interval CISI Program was developed in accordance with the requirements of 10 CFR 50.55a and the 2007 Edition with the 2008 Addenda of ASME Section XI, subject to the limitations and modifications contained within paragraph (b) of the regulation. With the update to the WBN Unit 1 CISI Program for the Third Interval, the WBN Unit 2 CISI Program was updated to the 2013 Edition of ASME Section XI through use of a request for alternative in accordance with 10 CFR 50.55a(g)(4)(iv). This update allowed WBN to utilize a single CISI Program.The WBN Unit 1 Third Interval is effective from September 9, 2018 through September 8th, 2028. The WBN Unit 2 First Interval is effective October 19th, 2016 through October 18th, 2026.Component Accessibility ISI Class MC and CC components subject to examination shall remain accessible for either direct or remote visual examination, from at least one side, per the requirements of ASME Section XI, Paragraph IWE-1230.Paragraph IWE-1231(a)(3) requires 80% of the pressure-retaining boundary that was accessible after construction to remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant.CNL-20-006 E1-13 of 41

Enclosure 1 Portions of components embedded in concrete or otherwise made inaccessible during construction are exempted from examination, provided that the requirements of ASME Section XI, Paragraph IWE-1232 have been fully satisfied.In addition, inaccessible surface areas exempted from examination include those surface areas where visual access by line of sight with adequate lighting from permanent vantage points is obstructed by permanent plant structures, equipment, or components; provided these surface areas do not require examination in accordance with the inspection plan, or augmented examination in accordance with paragraph IWE-1240. The areas around inaccessible areas, including those inaccessible due to the ice condenser configuration, are examined and, to date, no areas have been identified that would indicate there is an issue that would adversely affect structural integrity or leak tightness of the SCV in inaccessible areas.Responsible Individual The qualified primary CISI program owner serves as the Responsible Individual (RI) for containment inspections per IWE-2320. The specific responsibilities of the RI are provided below:

1. Development of plans and procedures for general visual examination of containment surfaces.

2. Instruction, training, and approval of general visual examination personnel.

3. Performance or direction of general visual examinations.

4. Evaluation of general visual examination results and documentation.

The CISI Program owner may delegate performance of these responsibilities to qualified individuals in the Inspection Services Organization, but retains overall responsibility for their implementation.Owner Elected Examinations The WBN Unit 1 Steam Generators were replaced during U1R7. The inside surface of the two SCV cut-outs in the top of the containment dome, which were removed for steam generator replacement, remain uncoated and shall be monitored for wall thickness due to potential corrosion activity. These areas do not have the potential for accelerated corrosion activity, and any corrosion is expected to be generally uniform across the uncoated surface. Any degradation is expected to occur in the area surrounding the cut line weld toe. Examinations were initially performed every other outage beginning U1R8. After six cycles of operation, the examination areas have shown no measurable corrosion. Trending of the examination results showed corrosion of the liner equal to 10% wall loss would not occur for approximately 125 years in one grid location and more then 300 years in all other locations. Based on these trending results, the examination frequency has been extended to once every 10 years. Examinations consist of performing a 100% ultrasonic thickness examination to the extent possible (except for weld surface) of three grids per cut line (two cuts). Each grid is to be 12 inches wide (six inches on each side from the center on the weld) by 24 inches long.There are currently no owner elected examinations for WBN Unit 2 containment.CNL-20-006 E1-14 of 41

Enclosure 1 Table 3.4.1-1 Components Subject to Examination Unit 1 EXAM ITEM DESCRIPTION EXAM NUMBER INTERVAL NUMBER FREQ. OF NUMBER NUMBER NUMBER DRAWING CAT. NO. METHOD OF REQMNT. EXAMS EXAM / OF EXAMS OF EXAMS OF EXAMS NO.COMPONENTS FOR FOR DEFERRAL 1st 2nd 3rd EXAM INTERVAL PERIOD PERIOD PERIOD E-A CONTAINMENT SURFACES E-A E1.10 Containment Vessel Pressure Retaining Boundary E-A E1.11 Accessible General 1 300% 3 Each 1 1 1 ISI-503-C Surface Areas Visual Inspection Series (Note 1) Period E-A E1.12 Accessible Visual 0 100% 0 End of 0 0 0 Surface Areas Interval VT-3 E-A E1.20 Vent System - Visual 0 100% 0 End of 0 0 0 Accessible Interval Surface Areas VT-3 E-A E1.30 Moisture General 1 Thermal 300% 57 Each 19 19 19 ISI-503-C Barriers Visual Barrier Inspection Series Period 18 LCC Boxes E-C CONTAINMENT SURFACES REQUIRING AUGMENTED EXAMINATION E-C E4.10 Containment Surface Areas E-C E4.11 Visible Surfaces Visual Unit 1 Category E-C Augmented Examinations Schedule VT-1 (None Currently Identified)E-C E4.12 Surface Area Volumetric Unit 1 Category E-C Augmented Examinations Schedule Grid, Minimum.Wall Thickness UT (None Currently Identified)Location CNL-20-006 E1-15 of 41

Enclosure 1 Components Subject to Examination Unit 1 EXAM ITEM DESCRIPTION EXAM NUMBER INTERVAL NUMBER FREQ. OF NUMBER NUMBER NUMBER DRAWING CAT. NO. METHOD OF REQMNT. EXAMS EXAM / OF EXAMS OF EXAMS OF EXAMS NO.COMPONENTS FOR FOR DEFERRAL 1st 2nd 3rd EXAM INTERVAL PERIOD PERIOD PERIOD E-G Pressure Retaining Bolting E-G E8.10 Bolted Visual 34 100% 34 Each 0 0 34 ISI-503-C Connections Interval Series VT-1 (Note 2)APP- Owner Elected Examinations DAPP- 2.0 SCV Dome UT 6 6 6 Each 0 0 6 ISI-503-C D Interval Series Notes:

1. The 1 area, which is the entire SCV, is divided into ten examinations, five on the interior surface and five on the exterior surface of the SCV. The five interior/exterior exams are the four 90-degree quadrants of the vertical cylinder and the dome of the SCV.

2. Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled.

CNL-20-006 E1-16 of 41

Enclosure 1 Table 3.4.1-2 Components Subject to Examination Unit 2 EXAM ITEM DESCRIPTION EXAM NUMBER INTERVAL NUMBER FREQ. OF NUMBER NUMBER NUMBER DRAWING CAT. NO. METHOD OF REQMNT. EXAMS EXAM / OF EXAMS OF EXAMS OF EXAMS NO.COMPONENTS FOR FOR DEFERRAL 1st 2nd 3 rd EXAM INTERVAL PERIOD PERIOD PERIOD E-A CONTAINMENT SURFACES E-A E1.10 Containment Vessel Pressure Retaining Boundary E-A E1.11 Accessible General 1 300% 3 Each 1 1 1 ISI-20MC-E Surface Areas Visual Inspection Series (See Note 1) Period E-A E1.12 Accessible Visual 0 100% 0 End of 0 0 0 Surface Areas Interval VT-3 E-A E1.20 Vent System - Visual 0 100% 0 End of 0 0 0 Accessible Interval Surface Areas VT-3 E-A E1.30 Moisture General 1 Thermal 300% 33 Each 11 11 11 ISI-20MC-E Barriers Visual Barrier Inspection Series Period 10 LCC Boxes E-C CONTAINMENT SURFACES REQUIRING AUGMENTED EXAMINATION E-C E4.10 Containment Surface Areas E-C E4.11 Visible Surfaces Visual Unit 2 Category E-C Augmented Examinations Schedule VT-1 (None Currently Identified)E-C E4.12 Surface Area Volumetric Unit 2 Category E-C Augmented Examinations Schedule Grid, Minimum Wall Thickness UT (None Currently Identified)Location CNL-20-006 E1-17 of 41

Enclosure 1 Components Subject to Examination Unit 2 EXAM ITEM DESCRIPTION EXAM NUMBER INTERVAL NUMBER FREQ. OF NUMBER NUMBER NUMBER DRAWING CAT. NO. METHOD OF REQMNT. EXAMS EXAM / OF EXAMS OF EXAMS OF EXAMS NO.COMPONENTS FOR FOR DEFERRAL 1st 2nd 3 rd EXAM INTERVAL PERIOD PERIOD PERIOD E-G Pressure Retaining Bolting E-G E8.10 Bolted Visual 38 100% 38 Each 0 0 38 ISI-20MC-E Connections Interval Series VT-1 (See Note 2)Notes:

CNL-20-006 E1-18 of 41

Enclosure 1 General Visual Examinations A General Visual examination shall be performed once each period per Table IWE-2500-1.Examinations include all accessible interior and exterior surfaces of Class MC components, parts, and appurtenances. The following items are examined during the general visual examinations:

1. Integral attachments and structures that are parts of reinforcing structure, such as stiffening rings, manhole frames, and reinforcement around openings.

2. Surfaces of attachment welds between structural attachments and the pressure retaining boundary or reinforcing structure, except for nonstructural or temporary attachments as defined in NE-4435.

3. Surfaces of containment structural and pressure boundary welds, including longitudinal welds (Category A), circumferential welds (Category B), flange welds (Category C), and nozzle-to-shell welds (Category D) as defined in NE 33511 for Class MC; and surfaces of Flued Head and Bellows Seal Circumferential Welds joined to the Penetration.

General Visual Examination History Unit 1 The most recent category E-A, Item E1.11 examinations performed on Unit 1 were conducted in accordance with TVA NDE Procedure N-VT-25 during U1R15. There were no relevant indications that required evaluation. The majority of indications documented were chipped or peeling paint and some light surface rust in some areas.Unit 2 The most recent category E-A, Item E1.11 examinations performed on Unit 2 were conducted in accordance with TVA NDE Procedure N-VT-25 during U2R2. These examinations were completed prior to the performance of the CILRT. The majority of indications documented were chipped or peeling paint and some light surface rust in some areas.Moisture Barrier Examinations The moisture barrier examinations include moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application.Thermal Barrier Flashing The WBN design does not include a moisture barrier seal as depicted in Figure IWE-2500-1. However, the Thermal Barrier Flashing, which was installed to protect the thermal insulation attached to the SCV for post-accident conditions when high-temperature water may accumulate upon containment floor, inadvertently acts as a moisture barrier. The flashing extends approximately 4-4 above the containment raceway floor (El. 702) and is sealed against the SCV with sealant to minimize water intrusion under accident conditions.CNL-20-006 E1-19 of 41

Enclosure 1 Leak Chase Channel Boxes The WBN design contains Leak Chase Channel (LCC) Boxes in the raceway on El. 702' and the keyway on El. 674'. Each LCC box consists of a closure lid that is sealed with a rubber gasket and/or sealant. The scope of the LCC box moisture barrier examination is limited to the LCC box cover gasket and/or sealant. Inside each LCC box is a pipe stub that leads to the LCC system below. Each pipe stub is equipped with a threaded pipe cap, but is not part of the moisture barrier.Moisture Barrier Examination History Thermal Barrier The Thermal Barrier examinations are normally performed at the end of the refueling outage after the Raceway area has been cleared of outage related equipment and material. The Raceway area experiences high traffic during outages and gouging or denting of the thermal Barrier Flashing is a common observation.Unit 1 The most recent Thermal Barrier, category E-A, Item E1.30 examinations performed on Unit 1 were conducted in accordance with TVA NDE Procedure N-VT-16 during U1R14.Observations: Numerous dents and gouges were observed on the metal flashing along the entire circumference. No breach or puncture in the metal flashing was identified. The moisture seal barrier was observed not to be separated or cracked.No leak paths noted.Unit 2 The most recent Thermal Barrier, category E-A, Item E1.30 examinations performed on Unit 2 were conducted in accordance with TVA NDE Procedure N-VT-25 during U2R2.Observations: Some rivets were broken/missing on the first horizontal seam from the top. Some vertical seams were loose/bulging. The first horizontal seam from the floor was not sealed. In all occurrences the damage was indicative of damage related to outage activities within the Raceway. The damage was not indicative of service induced flaws nor was there any indication of moisture intrusion. There was no evaluation required as all areas had corrective measures performed to the extent necessary to meet the acceptance standards of IWE-3512. Because no areas of the flashing were indicative of degradation in an inaccessible area, there is no engineering evaluation required in accordance with IWE-2500(d).Leak Chase Channel Boxes Unit 1 The most recent LCC Box, category E-A, Item E1.30 examinations performed on Unit 1 were conducted in accordance with TVA NDE Procedure N-VT-16 during U1R14. During performance of these examinations, water was discovered within seven of the LCC Boxes indicating the moisture barrier material had failed. Water samples were removed from the boxes for a chemistry analysis and the remaining water was subsequently removed. Once all of the water was removed a boroscope was used to assess the inaccessible area of the bottom liner plate below the LCC Box. The visual examination concluded there was no accelerated corrosion of the CNL-20-006 E1-20 of 41

Enclosure 1 bottom liner plate occurring. Chemical analysis concluded that the water was not ground water intrusion. The source of water had some minor boron concentration present, but was not consistent with the boron concentration of primary water. The moisture barrier materials were replaced and all LCC boxes were re-sealed.Unit 2 The most recent category E-A, Item E1.30 examinations performed on Unit 2 were conducted in accordance with TVA NDE Procedure N-VT-25 during U2R2. During performance of these examinations water was discovered within three of the LCC Boxes indicating the moisture barrier material had failed. Water samples were removed from the boxes for a chemistry analysis and the remaining water was subsequently removed. Once all of the water was removed a boroscope was used to assess the inaccessible area of the bottom liner plate below the LCC Box. The visual examination concluded there was no accelerated corrosion of the bottom liner plate occurring. The chemical analysis concluded that the water was not ground water intrusion. The source of water had some minor boron concentration present, but was not consistent with the boron concentration of primary water. The moisture barrier materials were replaced and all LCC boxes were re-sealed.3.4.2 Containment Coatings Inspections The WBN Protective Coatings Program is designed to install and maintain all Coating Service Level (CSL) I, II, III, and corrosive environment protective coatings at the quality required to perform their intended function. These inspections assure conformance to the Watts Bar Nuclear Plant response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors."Results of Recent Coatings Inspections Unit 1 Inspections of the coatings in the Unit 1 reactor containment building were completed in U1R15 in the fall of 2018 and assessed by Engineering. The assessment was compared with a previous containment coating assessment from U1R14 in accordance with license basis requirements and coating program procedures. Overall, no significant findings were identified. The condition of the containment coatings was acceptable and no immediate corrective actions were required to meet design and license basis requirements. Some minor findings were noted and a condition report was generated. The total quantity of unqualified coatings remains within the bounds required by design basis, with considerable margin remaining.Unit 2 Inspections of the coatings in the Unit 2 reactor containment building were completed in U2R2 (spring of 2019) and assessed by Engineering. The assessment was compared with a previous containment coating assessment from U2R1 in accordance with license basis requirements and coating program procedures.Overall, no significant findings were identified. The condition of the containment coatings was acceptable and no immediate corrective actions were required to meet design and license basis requirements. Some minor findings were noted and a condition report was generated. The total quantity of unqualified coatings remains within the bounds required by design basis, with considerable margin remaining.CNL-20-006 E1-21 of 41

Enclosure 1 3.5 Industry Operating Experience Review The NRC has issued several information notices and a regulatory issue summary concerning containment corrosion. TVA reviewed these notices to determine the impact on the WBN Unit 1 and Unit 2 containments.Information Notice (IN) 1992-20, "Inadequate Local Leak Rate Testing," dated March 3, 1992, stated that problems exist with testing of stainless steel containment penetration bellows. If the plies of the bellows are in contact with each other, the flow of the test medium would be restricted and in-leakage through such bellows may not be readily detectable by LLRTs. The bellows used at WBN contain a full wire mesh between the plies to maintain a uniform gap.Therefore, this problem would not occur at WBN.IN 2004-09, Corrosion of Steel Containment and Containment Liner, dated April 27, 2004, addresses concerns identified by the NRC for corrosion in freestanding metallic containments and in liner plates of reinforced and pre-stressed concrete containments. Specifically, the integrity of the moisture barrier seal at the floor-to-liner or floor-to-containment junction was noted as important for avoiding conditions favorable to corrosion and thinning of the containment liner plate material. In response to the situations identified in this IN, WBN containment inspection procedures were revised to provide special attention for the SCV to concrete interfaces.IN 2010-12, Containment Liner Corrosion, dated June 18, 2010, was issued to inform addressees of recent issues involving corrosion of the steel reactor containment building liner. The examples dealt with containment liner corrosion resulting from liner plates in contact with objects and materials that are lodged between or embedded in the containment concrete, including organic materials.WBN containment inspection procedures were reviewed with respect to this issue and were found to be adequate for detecting such situations.IN 2011-15, "Steel Containment Degradation and Associated License Renewal Aging Management Issues," dated August 1, 2011, describes mechanisms that can lead to degradation of coatings and pitting of containment liner plates due to long term exposure to water and moisture. IN-2011-15 was reviewed for applicability to WBN. With respect to the mechanisms described in this IN, the WBN SCV is monitored regularly using walkdowns, with input provided to the Maintenance Rule Program.IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner," dated May 5, 2014, provided examples of operating experience at some plants of water accumulation and corrosion degradation in the leak-chase channel system that has the potential to affect the leak-tight integrity of the containment shell or liner plate. In response to the situations identified in this IN, WBN containment inspection procedures were revised to incorporate moisture barrier examination requirements for the leak chase channel boxes.Regulatory Issue Summary 2016-07, Containment Shell or Liner Moisture Barrier Inspection, dated May 9, 2016, was issued to reiterate the NRC staffs position in regard to inservice inspection requirements for moisture barrier CNL-20-006 E1-22 of 41

Enclosure 1 materials as discussed in the ASME B&PV Code, Section XI, Subsection IWE.TVA determined that such materials were already included in the containment inspection procedures for WBN Unit 1. WBN Unit 2 inspection procedures were revised to include the thermal barrier flashing as a moisture barrier component under ASME Section XI, Examination Category E-A, E1.30.3.6 NRC Limitations and Conditions for NEI 94-01 3.6.1 June 25, 2008 NRC Safety Evaluation The limitations and conditions from the June 25, 2008 safety evaluation (Reference 8) for NEI 94-01, Revision 2 are presented in Table 3.6.1-1 with the TVA response.Table 3.6.1-1 Limitation/Condition Item (From Section 4.1 of SE dated TVA Response June 25, 2008 )

1. For calculating the Type A leakage TVA will utilize the definition in NEI 94-01, rate, the licensee should use the Revision 3-A, Section 5.0. This definition definition in the NEI TR 94-01, has remained unchanged from Revision Revision 2, in lieu of that in 2-A to Revision 3-A of NEI 94-01.

ANSI/ANS-56.8-2002.

2. The licensee submits a schedule of TVA will perform a general visual containment inspections to be inspection of the accessible interior and performed prior to and between Type exterior surfaces of the primary A tests. containment and components prior to the Type A test. Inspections performed between Type A tests are described further in Section 3.4 of this enclosure.

3. The licensee addresses the areas of TVA will perform general visual the containment structure potentially observations of the accessible interior and subjected to degradation. external surfaces of the containment structure in accordance with containment inspection procedures. Any evidence of structural deterioration is recorded and evaluated or repaired as required. These inspections are described further in Section 3.4 of this enclosure.

CNL-20-006 E1-23 of 41

Enclosure 1 Limitation/Condition Item (From Section 4.1 of SE dated TVA Response June 25, 2008 )

4. The licensee addresses any tests TVA will implement the staff position with and inspections performed following regard to any future post-repair pressure major modifications to the testing following major WBN, Unit 1 and 2, containment structure, as applicable. containment repairs and modifications, as explained in Section 3.1.4 of the NRC staff SE for NEI 94-01, Revision 2. Specifically, with regards to major repairs and modifications, TVA recognizes that, since the issuance of the SE for NEI 94-01 Revision 2, the requirement to perform a Type A test has been removed from ASME Section XI, Subsection IWE-5000 and is now NRC condition 10CFR50.55a(b)(2)(ix)(J). Following a major modification, as an alternative to performing a Type A test TVA will perform the following activities.

a) Perform all NDE required by the construction code.b) Examine the locally welded areas for essentially zero leakage using a bubble test or equivalent.c) Subject the entire containment to Pa pressure specified in Technical Specifications for a minimum of 10 minutes.d) Perform a general visual examination of the accessible portions of the interior and exterior surfaces of containment in accordance with ASME B&PV Code, Subsection IWE.

5. The normal Type A test interval TVA will comply with this condition.

should be less than 15 years. If a Extensions of the Type A test frequency licensee has to utilize the provision of beyond 15 years are used only to Section 9.1 of NEI TR 94-01, accommodate unforeseen emergent Revision 2, related to extending the conditions.ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.CNL-20-006 E1-24 of 41

Enclosure 1 Limitation/Condition Item (From Section 4.1 of SE dated TVA Response June 25, 2008 )

6. For plants licensed under Not applicable.

10 CFR Part 52, applications WBN Units 1 and 2 were not licensed requesting a permanent extension of pursuant to 10 CFR Part 52.the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.3.6.2 June 8, 2012 NRC Safety Evaluation The two conditions from Section 4.0 of the June 8, 2012 safety evaluation (Reference 14) for NEI 94-01, Revision 3 are stated below with the TVA response.Table 3.6.2-1 Limitation/Condition Item (From Section 4.0 of SE dated TVA Response June 8, 2012)Condition 1 presents three separate items that are required to be addressed.1a. The staff is allowing the extended The post-outage report will include the interval for Type C LLRTs be margin between the Type B and Type C increased to 75 months with the Minimum Pathway Leak Rate (MNPLR) requirement that a licensee's summation value, as adjusted to include post-outage report include the margin the estimate of applicable Type C leakage between the Type B and Type C understatement, and its regulatory limit of leakage rate summation and its 0.60La. TVA will establish an administrative regulatory limit. limit to provide margin to the regulatory limit of 0.60La.CNL-20-006 E1-25 of 41

Enclosure 1 Limitation/Condition Item (From Section 4.0 of SE dated TVA Response June 8, 2012) 1b. In addition, a corrective action plan When the potential leakage understatement shall be developed to restore the adjusted Types B and C MNPLR total is margin to an acceptable level. greater than the WBN administrative leakage summation limit, but less than the regulatory limit of 0.60La, then an analysis and determination of a corrective action plan will be prepared to restore the leakage summation margin to less than the WBN leakage limit. This plan will focus on those components which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action that best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.1c. Use of the allowed 9-month TVA will apply the 9-month extension extension for eligible Type C valves period only to eligible Type C components is only authorized for non-routine for non-routine emergent conditions. Such emergent conditions. At no time occurrences will be documented in the shall an extension be allowed for record of tests. This non-routine extension Type C valves that are restricted is not allowed for valves specifically categorically (e.g. BWR MSIVs), and restricted to a maximum 30 month interval those valves with a history of or any valve held to less than a maximum leakage, or any valves held to either interval or the base interval (30 months).a less than maximum interval or to the base refueling cycle interval.CNL-20-006 E1-26 of 41

Enclosure 1 Limitation/Condition Item (From Section 4.0 of SE dated TVA Response June 8, 2012)Condition 2 presents two separate items that are required to be addressed.2a. Extending the LLRT intervals TVA will apply a potential leakage beyond 60-months to a 75-month understatement adjustment factor to the interval should be similarly actual As-Left leak rate. This will result in a conservative provided an estimate is combined conservative Type C total for all made of the potential understatement 75-month LLRTs being "carried forward" and its acceptability determined as and will be included when the total leakage part of the trending specified in summation is required to be updated (either NEI TR 94-01, Revision 3, while on line or following an outage).Section 12.1.2b. When routinely scheduling any LLRT A post-outage report is prepared with valve interval beyond 60-months and results of the tests performed during that up to 75-months, the primary outage. The report will show that the containment leakage rate testing applicable performance criteria are met and program trending or monitoring must serve as a record that continuing include an estimate of the amount of performance is acceptable. If an adverse understatement in the Types B and C trend in the potential leakage total and must be included in a understatement is identified, then a licensee's post-outage report. The corrective action plan is prepared, focused report must include the reasoning on those components which have and determination of the acceptability contributed the most to the adverse trend in of the extension, demonstrating that the leakage summation value.the LLRT totals calculated represent the actual leakage potential of the penetrations.In addition, NEI 94-01, Revision 3-A also has a margin related requirement as contained in Section 12.1, Report Requirements.A post-outage report shall be prepared At WBN, in the event an adverse trend presenting results of the previous cycles Type B in the aforementioned potential and Type C tests, and Type A, Type B and leakage understatement adjusted Type Type C tests, if performed during that outage. B and C summation is identified, then The technical contents of the report are an analysis and determination of a generally described in ANSI/ANS-56.8-2002 and corrective action plan will be prepared shall be available on-site for NRC review. The to restore the trend and associated report shall show that the applicable margin to an acceptable level. The performance criteria are met and serve as a corrective action plan will focus on record that continuing performance is acceptable. those components which have The report shall also include the combined contributed the most to the adverse Type B and Type C leakage summation, and the trend in the leakage summation value margin between the Type B and Type C leakage and the manner of timely corrective rate summation and its regulatory limit. Adverse action, as deemed appropriate that trends in the Type B and Type C leakage rate best focuses on the prevention of summation shall be identified in the report and a future component leakage corrective action plan developed to restore the performance issues.margin to an acceptable level.CNL-20-006 E1-27 of 41

Enclosure 1 3.7 Plant-Specific Confirmatory Analysis 3.7.1 Methodology An evaluation has been performed to assess the risk impact of extending the WBN CILRT Type A interval from the current licensing basis (CLB) of once per ten years to a proposed licensing basis (PLB) of once per fifteen years. This evaluation is provided as Enclosure 2 to this LAR.A simplified bounding analysis consistent with the Electric Power Research Institute (EPRI) approach was used for evaluating the change in risk associated with increasing the test interval to fifteen years. The approach is consistent with that presented in:Appendix H of Electric Power Research Institute, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325,"EPRI Topical Report TR-1018243, October 2008 (Reference 12)Electric Power Research Institute, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," EPRI Topical Report TR-104285, August 1994 (Reference 6)Nuclear Regulatory Commission, "Performance-Based Containment Leak-Test Program," NUREG-1493, September 1995 (Reference 5)Calvert Cliffs liner corrosion analysis described in a letter to the NRC dated March 27, 2002 (Reference 16)The analysis uses results from the WBN analysis of core damage scenarios (Level 1) and subsequent containment responses (Level 2) resulting in various fission product release categories (including intact containment or negligible release). The applicable figures of merit for this risk-informed application include:Total LERF (Large Early Release Frequency) - Acceptance Criteria from RG 1.174 (Reference 11) <1.0E-05/yr (All Hazards)Change in LERF - Acceptance Criteria from RG 1.174, <1.0E-06/yr (All hazards)Population Dose Rate - Acceptance Criteria EPRI 1018243 §2.2 (Reference 12)

 <1.0 person-rem/yr or, Percent Increase in Population Dose - Acceptance Criteria EPRI 1018243 §2.2, <1.0% of total dose increase whichever is less restrictive Condition Containment Failure Probability (CCFP) - Acceptance Criteria EPRI 1018243 §2.2, Less than or equal to 1.5% increase In the safety evaluation issued by NRC letter dated June 25, 2008 (Reference 8), the NRC concluded that the methodology in EPRI Report No. 1009325, Revision 2 (Reference 10) is acceptable for referencing by licensees proposing to amend their TS to permanently extend the Type A surveillance test interval to 15 years, subject to the conditions noted in Section 4.2 of the safety evaluation.

CNL-20-006 E1-28 of 41

Enclosure 1 Table 3.7.1-1 addresses each of the four conditions for the use of EPRI Report No. 1009325, Revision 2 (from Section 4.2 of NRC Safety Evaluation dated June 25, 2008) (Reference 8).Table 3.7.1-1 Limitation/Condition Item (From Section 4.2 of SE dated TVA Response June 25, 2008)

1. The licensee submits documentation WBN PRA technical adequacy is indicating that the technical adequacy addressed in Section 4.0 of the PRA of their PRA (Probabilistic Risk Evaluation. (Reference 17)

Assessment) is consistent with the requirements of RG 1.200 relevant to the CILRT extension application.2a. The licensee submits documentation The containment Type A ILRT does not indicating that the estimated risk mitigate or support the mitigation of core increase associated with permanently damage; however, it does identify extending the CILRT surveillance potential leakage paths from within interval to 15 years is small, and containment to the environment. The consistent with the clarification relevant figure of merit is the large early provided in Section 3.2.4.5 of this SE. release frequency (LERF). Using the methodology from the EPRI guidance, the increase in LERF resulting from a change in the Type A ILRT test interval from three tests-in-ten years to one test-in-fifteen years is estimated as 1.90E-07/year for both units (internal events and external events). The total LERF is 1.61E-6/year for Unit 1 and 1.60E-6/year for Unit 2. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174.2b. Specifically, a small increase in The postulated impact on the population population dose should be defined as dose and dose-rate resulting from an increase in population dose of less changing the Type A test frequency to the than or equal to either 1.0 person-rem proposed licensing basis from the original per year or 1% of the total population licensing basis (OLB) of three tests-in-ten dose, whichever is less restrictive. years to one test-per-fifteen years is determined using the method discussed in Reference 12.The impact is measured as an increase to the total integrated plant dose for those accident sequences influenced by Type A testing, is 0.058 person-rem/year, or 0.4%(both units). NEI 94-01 states that a small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT CNL-20-006 E1-29 of 41

Enclosure 1 Limitation/Condition Item (From Section 4.2 of SE dated TVA Response June 25, 2008) intervals. The results of this evaluation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to severe accident risks is negligible.Section 7.3 of the PRA Evaluation in Reference 17 presents the population dose information.2c. In addition, a small increase in CCFP The increase in the conditional should be defined as a value containment failure probability (CCFP) marginally greater than that accepted from the three tests-in-ten years interval in a previous one-time 15-year ILRT to one test-in-fifteen years interval is extension requests. This would 0.908%. NEI 94-01 states that an require that the increase in CCFP be increase in CCFP of 1.5% is small.less than or equal to 1.5 percentage Therefore, this increase is judged to be point. small. Section 7.4 of the PRA evaluation in Reference 17 presents the conditional containment failure probability information.

3. The methodology in EPRI Report EPRI Class 3b represents a Large, No. 1009325, Revision 2, is pre-Existing Leak in the Containment acceptable except for the calculation liner. All core damage accident of the increase in expected population progression bins with a pre-existing leak dose (per year of reactor operation). in the containment structure in excess of In order to make the methodology normal leakage (La) are characterized as acceptable, the average leak rate for >10 La. For this evaluation, the the pre-existing containment large representative containment leakage for leak rate accident case (accident Class 3b sequences used by WBN is case 3b) used by the licensees shall 100 La, based on the guidance provided in be 100 La instead of 35 La. Reference 12.

4. A LAR is required in instances where WBN does not credit containment containment over-pressure is relied over-pressure for ECCS performance.

upon for ECCS performance.3.7.2 Probabilistic Risk Assessment (PRA) Acceptability The WBN Internal Events, Internal Flood, and Seismic PRA models have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models credited in this request are the PRA models used in the TSTF-425 Surveillance Frequency Control Program (SFCP) application (Reference 18), with routine maintenance updates applied. Capability Category (CC) II of the NRC-endorsed ASME/ANS PRA Standard is the target capability level for both of these applications. The acceptability (previously referred to as technical adequacy or quality) of the PRA models was reviewed by the NRC for that application and determined to be acceptable, as discussed in the Safety Evaluation (Reference 19).CNL-20-006 E1-30 of 41

Enclosure 1 The NRC Safety Evaluation also states the assessment of external events can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Therefore, the CILRT interval extension risk assessment is allowed to use the existing models and other existing fire and external hazard evaluations.Section 3.2 of the WBN TSTF-425 submittal (Reference 18) provides a more detailed discussion of the external hazard evaluations. The information in Section 2 of the WBN TSTF-425 submittal demonstrates that the PRA is of sufficient quality and level of detail to support this submittal, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.3.7.3 Conclusions of the Plant-Specific Risk Assessment Results The findings of the WBN risk assessment confirm the general findings of previous studies that the risk impact associated with extending the original licensing basis of Type A test intervals from three tests in ten years to one test in 15 years is small.The WBN plant-specific results for extending the Type A test interval from three-in-ten years to one-in-15 years is summarized in Table 3.7.3-1.The containment liner does not perform a core damage mitigation function; therefore, extending the CILRT interval has no effect on core damage frequency. Furthermore, WBN does not rely on containment overpressure to assure adequate net positive suction head is available for emergency core cooling system pumps taking suction from the containment sump following design basis accidents.Table 3.7.3-1 Unit 1 PRA Results Acceptable Metric Value Acceptance Criteria for Application?LERFIE-Total PLB 1.37E-06/yr

 <1.0E-05/rx-yr Yes LERFTotal(IE & EE) PLB 1.61E-06/yr LERFTotal(OLBCLB) EE&IE 1.11E-07/yr <1.0E-06/rx-yr Yes LERFTotal(OLBPLB) EE&IE 1.90E-07/yr CCFP(OLBCLB), Inc. Corrosion 0.529%

1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.908%DOSE(OLBCLB) 3.37E-02 per-rem/yr DOSE(OLBPLB) 5.75E-02 per-rem/yr <1.0 person-rem/yr or <1% of total dose, whichever is less Yes

 %DOSE(OLBCLB) 0.24% restrictive. %DOSE(OLBPLB) 0.41%

CNL-20-006 E1-31 of 41

Enclosure 1 Unit 2 PRA Results Acceptable Metric Value Acceptance Criteria for Application?LERFIE-Total PLB 1.36E-06

 <1.0E-05/rx-yr Yes LERFTotal(IE & EE) PLB 1.60E-06 LERFTotal(OLBCLB) EE&IE 1.11E-07 <1.0E-06/rx-yr Yes LERFTotal(OLBPLB) EE&IE 1.90E-07 CCFP(OLBCLB), Inc. Corrosion 0.529%

1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.908%DOSE(OLBCLB) 3.34E-02 per-rem/yr DOSE(OLBPLB) 5.76E-02 per-rem/yr <1.0 person-rem/yr or <1% of total dose, whichever is less Yes

 %DOSE(OLBCLB) 0.24% restrictive. %DOSE(OLBPLB) 0.42%

Based on the results in Table 3.7.3-1, the proposed 15-year Type A test interval represents a small change in risk and is acceptable as a permanent change. Details of the WBN risk assessments are contained in Enclosure 2 of the LAR.NEI 94-01, Revision 3-A, describes an NRC-accepted approach for implementing the performance-based requirements of Appendix J, Option B. It incorporates the regulatory positions stated in Regulatory Guide 1.163 and includes provisions for extending Type A test intervals to 15 years and Type C test intervals to 75 months.NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies.3.8 Basis for the Proposed TS Changes 3.8.1 General basis The provisions of Option B in Appendix J to 10 CFR Part 50 will continue to be met through compliance with:NEI 94-01 Revision 3-A, the Limitations & Conditions set forth in the NRC Safety Evaluations for NEI 94-01 Rev 3-A and Rev 2-A, and ANSI/ANS 56.8-2002.The adoption of NEI 94-01, Revision 3-A, is justified by the excellent performance history at WBN for Type A and Type C leakage test results and for containment inspections as described in Sections 3.3 and 3.4 of this LAR, as well as the risk insights provided by the plant-specific confirmatory analysis described in Section 3.7.TVA response to the Limitations & Conditions set forth in the NRC Safety Evaluations for NEI 94-01 Rev 3-A and Rev 2-A are described in Section 3.6.Compliance with ANSI/ANS 56.8-2002 is maintained through the use of procedures that already address the requirements and processes for implementation of NEI 94-01, Rev 3-A, at other plants within the TVA nuclear fleet.CNL-20-006 E1-32 of 41

Enclosure 1 3.8.2 Use of a bounding value for Pa TVA is requesting the use of a bounding value of 15.0 psig for Pa instead of the calculated Pa value as defined 10 CFR 50, Appendix J, Option B, Section II Definitions and ANSI/ANS 56.8, Section 2 Definitions. The current calculated Pa value is 9.36 psig for both Unit 1 and Unit 2 as shown in UFSAR Chapter 6.2.1.3.3 Long-Term Containment Pressure Analysis. A lower limit of 9.0 psig for the calculated Pa was used in evaluating acceptability for the bounding Pa of 15.0 psig.The purpose of requesting a bounding value for Pa is to minimize the impact on related documents when the calculated Pa is changed. For example, if the calculated Pa changes, then revisions of TS 5.2.7.19 and approximately 43 containment leak rate test procedures, among others, would be required. Use of a bounding value for Pa eliminates that burden and provides a similar level of quality and safety as described below.NEI 94-01, Revision 3-A, Section 8.0, states, in part, Type A, Type B, and Type C tests should be performed using the technical methods and techniques specified in ANSI/ANS-56.8-2002, or other alternative testing methods that have been approved by the NRC.ANSI/ANS 56.8-2002 provides the following definitions and requirements that are sensitive to the use of a higher bounding value for Pa.La (wt%/24 h): The maximum allowable Type A test leakage rate at Pa.Pa (psig or kPa): The calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA).Type B and Type C tests shall be conducted at a differential pressure of greater than or equal to Pa unless otherwise specified in the plant's licensing basis. When a higher differential pressure results in increased sealing, the differential pressure shall not exceed 1.1 Pa.In most cases, the use of a higher bounding Pa pressure is more conservative (e.g., results in a higher leak rate for components where higher pressure reduces sealing) than use of a lower calculated Pa pressure. However, there are two cases where use of a bounding value for Pa will result in a potential non-conservative deviation from ANSI/ANS-56.8-2002 as compared to testing using the lower calculated Pa.3.8.3 Deviation #1 from ANSI/ANS 56.8-2002 related to the use of a bounding Pa The first case is related to the allowable leakage rate (La). ANSI/ANSI 56.8-2002 defines La (wt%/24 h) as the maximum allowable Type A test leakage rate at Pa. It also defines Pa (psig or kPa) as the calculated peak containment internal pressure related to the design basis loss-of-coolant accident (LOCA). La is a mass leakage rate and there is a direct correlation with Pa pressure when converted to a volumetric leak rate. Therefore, a bounding Pa of 15.0 psig will result in a higher allowable leakage rate, La, than what would be permitted if using the calculated Pa value. The effect of Pa on La is illustrated by the simplified formula below. The resulting La at three different Pa values is shown in Table 3.8.3-1.CNL-20-006 E1-33 of 41

Enclosure 1 0.0025 (%wt/24hr) x Containment Volume (ft3) x [Pa (psig) + PATM (psia)]La

(scfh)PATM x 24 (hrs/day) 0.0025 x 1,171,012 x (Pa + 14.69595)La

(scfh) 14.69595 x 24 Table 3.8.3-1 Pa La 9.00 psig 196.68 scfh 9.36 psig 199.67 scfh 15.00 psig 246.48 scfh Historically, the WBN as-found minimum pathway total containment leak rate for Local Leak Rate Test (LLRT) components has been less than 14% of the 0.6 La TS acceptance criteria at 15.0 psig.The value for La at WBN has been based on a Pa value of 15.0 psig since initial startup and commercial operation. As such, the associated dose analysis has also used an La based on Pa of 15.0 psig as described in TS Bases B3.6.1, Applicable Safety Analysis, which states in part as follows.The containment was designed with an allowable leakage rate of 0.25% of containment air weight per day. This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J, Option B, as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) related to the design basis LOCA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing.La is assumed to be 0.25% per day in the safety analysis at Pa = 15.0 psig which bounds the calculated peak containment internal pressure resulting from the limiting design basis LOCA.Thus, no adverse effects are introduced by the use of this bounding Pa of 15.0 psig.3.8.4 Deviation #2 from ANSI/ANS 56.8-2002 related to the use of a bounding Pa The second case is related to the maximum test pressure. ANSI/ANS 56.8-2002 limits the maximum Type B and Type C test pressure to 1.1 times Pa for those components where a higher differential pressure results in increased sealing. This restriction is generically worded to apply to a broad range of component designs and is intended to prevent the use of a higher test pressure as a means to reduce the component leakage rate.To evaluate the effect of this difference in maximum test pressure, TVA reviewed all WBN components that are Type B and Type C tested, to identify the scope of components where a higher differential pressure may increase sealing. This subset scope of components was then evaluated in more detail. The purpose of the detailed evaluation was to determine whether an increase in seat leakage is CNL-20-006 E1-34 of 41

Enclosure 1 expected when the differential pressure (LLRT test pressure) is reduced from 16.5 psig to 9.0 psig. The maximum LLRT test pressure of 16.5 psig equates to 1.1 times the historical Pa value of 15.0 psig, which is the current and historical maximum LLRT test pressure allowed by WBN specific LLRT procedures. The lower limit of 9.0 psig for the evaluation was chosen to provide some margin below the current calculated Pa of 9.36 psig in case of future changes in the calculations.TVA contracted with Kalsi Engineering to perform this detailed evaluation. The full report of the Kalsi evaluation methodology, assumptions, and conclusion is provided in Enclosure 3 (Proprietary) and Enclosure 4 (Non-Proprietary).The detailed evaluation determined that an increase in seat leakage is not expected when the LLRT test pressure is reduced from 16.5 psig to a bounding lower limit of 9.0 psig. TVA plans to perform a confirmatory test, as recommended in the Kalsi Engineering report, during the WBN Unit 2 Cycle 3 Refueling Outage (U2R3) which is scheduled to commence in October 2020.3.9 Conclusion NEI 94-01, Revision 3-A, describes an NRC-accepted approach for implementing the performance-based requirements of Appendix J, Option B. It incorporates the regulatory positions stated in Regulatory Guide 1.163 and includes provisions for extending Type A test intervals to 15 years and Type C test intervals to 75 months.NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies.Based on the previous Type A tests conducted at WBN extension of the containment Type A test interval from ten to fifteen years represents minimal risk to increased leakage. The risk is further minimized by continued Type B and Type C testing performed in accordance with Appendix J, Option B, and the overlapping inspection activities performed as part of the following WBN inspection programs:Containment Inservice Inspection Program Protective Coatings Program This experience is supplemented by risk analysis studies, including the WBN risk analysis provided in Enclosure 2. The findings of the risk assessment confirm the general findings of previous industry studies, on a plant-specific basis, that extending the Type A test interval from ten to fifteen years results in a very small and acceptable change to the WBN baseline risk.CNL-20-006 E1-35 of 41

Enclosure 1

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria 4.1.1 Regulations 10 CFR 50.54(o), Conditions of licenses, requires that primary reactor containments for water cooled power reactors, other than facilities for which the certifications required under §§ 50.82(a)(1) or 52.110(a)(1) of this chapter have been submitted, shall be subject to the requirements set forth in appendix J to this part.10 CFR 50.54 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B identifies the performance-based requirements and criteria for preoperational and subsequent periodic leakage-rate testing.4.1.2 General Design Criteria General Design Criteria WBN Units 1 and 2 were designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN construction permit was issued in January 1973. The dual-unit UFSAR, however, addresses the NRC General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971, including Criterion 4 as amended October 27, 1987.GDC 16, Containment design, requires that reactor containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. Compliance with GDC 16 is described in Section 3.1.2.2 of the WBN UFSAR.GDC 50, Containment design basis, requires that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by § 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters. Compliance with GDC 50 is described in Section 3.1.2.5 of the WBN UFSAR.CNL-20-006 E1-36 of 41

Enclosure 1 GDC 52, Capability for containment leakage rate testing, requires that the reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure. Compliance with GDC 52 is described in Section 3.1.2.5 of the WBN UFSAR.GDC 53, Provisions for containment testing and inspection, requires that the reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows. Compliance with GDC 53 is described in Section 3.1.2.5 of the WBN UFSAR.With the implementation of the proposed changes, WBN Units 1 and 2 continue to meet the applicable regulations and requirements, subject to the previously approved exceptions.4.2 Precedent The following precedents are related to the proposed TS change in this submittal.Containment Leakage Rate Test Pressure The changes proposed in this LAR relative to Containment Leakage Rate Test Pressure are consistent with similar changes approved by the NRC for other nuclear power plants. These include changes approved for:Millstone Power Station, Unit 2, by License Amendment (LA) No. 326 dated March 31, 2016 (ML16068A312)Donald C. Cook Nuclear Plant, Units 1 and 2, by LA Nos. 336 and 318, respectively, dated June 7, 2017 (ML17131A277)Test Intervals The changes proposed in this LAR relative to Test Intervals are consistent with similar changes approved by the NRC for other nuclear power plants. These include changes approved for:Sequoyah Nuclear Plant, Units 1 and 2, by LA Nos. 335 and 328, respectively, dated November 30, 2015 (ML15320A218)Browns Ferry Nuclear Plant, Units 1, 2, and 3, by LA Nos. 305, 328, and 288, respectively, dated September 27, 2018 (ML18251A003)Cooper Nuclear Station by LA No. 266 dated August 18, 2020 (ML20191A273)Davis Besse Nuclear Power Station, Unit 1, by LA No. 300 dated August 24, 2020 (ML20213C726) 4.3 No Significant Hazards Consideration Tennessee Valley Authority (TVA) is requesting an amendment to Facility Operating Licenses NPF-90 and NPF-96 for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, respectively. The proposed amendment revises WBN Units 1 and 2 Technical Specification (TS) 5.7.2.19, "Containment Leakage Rate Testing Program," by adopting Nuclear Energy Institute (NEI) 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"(Reference 1) as the implementation document for the performance-based Option B CNL-20-006 E1-37 of 41

Enclosure 1 of 10 CFR Part 50, Appendix J. The proposed changes permanently extend the Type A containment integrated leak rate testing (CILRT) interval from 10 years to 15 years and the Type C local leakage rate testing intervals from 60 months to 75 months. In addition, a clarification of the value of Pa to be used for containment leakage rate testing purposes is incorporated.TVA evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," for development of the WBN containment leak rate program. NEI 94-01 allows, based on risk and performance, an extension of Type A and Type C containment leak test intervals. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses.The findings of the WBN risk assessment confirm the general findings of previous studies that the risk impact with extending the containment leak rate is small. Per the guidance provided in Regulatory Guide 1.174, an extension of the leak test interval in accordance with NEI 94-01, Revision 3-A results in an estimated change within the very small change region.Because the change is implementing a performance-based containment testing program, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled.The requirement for containment leakage rate acceptance will not be changed by this amendment. Therefore, the containment will continue to perform its design function as a barrier to fission product releases. The clarification of the value of Pa to be used for containment leakage rate testing purposes also does not involve a physical change to the plant nor a change in the manner in which the plant is operated or controlled Based on the above, it is concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not change the design or operation of structures, systems, or components of the CNL-20-006 E1-38 of 41

Enclosure 1 plant. The proposed change would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by this change. The clarification of the value of Pa to be used for containment leakage rate testing purposes also does not change the design or operation of structures, systems, or components of the plant.Based on the above, it is concluded that the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test and local leak rate testing frequency, does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation. The clarification of the value of Pa to be used for containment leakage rate testing purposes also has no effect on safety limits, limiting safety system setpoints, or limiting conditions for operation.The specific requirements and conditions of the TS Containment Leakage Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. This ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met with the acceptance of this proposed change since these are not affected by implementation of a performance-based containment testing program.Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the above, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),and, accordingly, a finding of no significant hazards consideration is justified.4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the CNL-20-006 E1-39 of 41

Enclosure 1 amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any radioactive effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Nuclear Energy Institute (NEI) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012 (ML12221A202)

2. Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), "Domestic Licensing of Production and Utilization Facilities," Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Reactors."

3. NRC Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 (ML003740058)

4. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, dated July 26, 1995 (ML11327A025)

5. NUREG-1493, "Performance-Based Containment Leak-Test Program," dated September 1995

6. EPRI Report 104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994

7. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2, dated August 2007 (ML072970206)

8. NRC letter to NEI, "Final Safety Evaluation For Nuclear Energy Institute (NEI)

Topical Report (TR) 94-01, Revision 2, "Industry Guideline For Implementing Performance-Based Option of 10 CFR PART 50, Appendix J" and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (TAC No. MC9663), dated June 25, 2008 (ML081140105)

9. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2-A, dated October 2008 (ML100620847)

10. EPRI Report 1009325, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," Revision 2, dated August 2007 (ML072970208)

CNL-20-006 E1-40 of 41

Enclosure 1

11. NRC Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018 (ML17317A256)

12. EPRI Report 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325," dated October 2008

13. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3, dated June 2011 (ML112920567)

14. NRC letter to NEI, "Final Safety Evaluation of Nuclear Energy Institute (NEI)

Report, 94-01, Revision 3, "Industry Guideline For Implementing Performance-Based Option of 10 CFR PART 50, Appendix J" (TAC No. ME2164)," dated June 8, 2012 (ML121030286)

15. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Revision 2, dated March 2009 (ML090410014)

16. Constellation Nuclear letter to NRC, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, dated March 27, 2002 (ML020920100)

17. PRA Evaluation WBN-0-19-078, WBN CILRT LAR PRA (provided as Enclosure 2)

18. TVA Letter to NRC, CNL-18-067, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (WBN-TS-18-14),

dated October 12, 2018 (ML18288A352)

19. NRC letter to TVA, Watts Bar Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 132 and 36 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-425, Revision 3, dated February 28, 2020 (ML20028F733)

CNL-20-006 E1-41 of 41

Enclosure 1 Attachment 1 Proposed TS Changes (Mark-Ups) for WBN Unit 1 CNL-20-006

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.18 Safety Function Determination Program (SFDP) (continued)A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 15.0 psig.For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound a range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.(continued)Watts Bar-Unit 1 5.0-24 Amendment 5, 63, 135, XX

Enclosure 1 Attachment 2 Proposed TS Changes (Mark-Ups) for WBN Unit 2 CNL-20-006

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J,"Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound the a range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.The maximum allowable containment leakage rate, La, at Pa, is 0.25%of the primary containment air weight per day.(continued)Watts Bar - Unit 2 5.0-25 Amendment 11, 39 Amendment XX

Enclosure 1 Attachment 3 Proposed TS Changes (Final Typed) for WBN Unit 1 CNL-20-006

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound a range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.(continued)Watts Bar-Unit 1 5.0-24 Amendment 5, 63, 135, XX

Enclosure 1 Attachment 4 Proposed TS Changes (Final Typed) for WBN Unit 2 CNL-20-006

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, as modified below:For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound a range of peak calculated containment internal pressures from 9.0 to 15.0 psig for the design basis loss of coolant accident.The maximum allowable containment leakage rate, La, at Pa, is 0.25%of the primary containment air weight per day.(continued)Watts Bar - Unit 2 5.0-25 Amendment 11, 39 Amendment XX

Enclosure 1 Attachment 5 Proposed TS Bases Page Changes (Mark-Ups) for WBN Unit 1 (For Information Only)CNL-20-006

Containment B 3.6.1 BASES SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. Failure to meet air lock, Shield Building containment bypass leakage path, and purge valve with resilient seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. As left leakage prior to the first startup after performing a required leakage test is required to be < 0.6 La for combined Type B and C leakage and d 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of d 1.0 La. At d 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance-Based Requirements."

2. Watts Bar FSAR, Section 15.0, "Accident Analysis."

3. Watts Bar FSAR, Section 6.2, "Containment Systems."

4. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995.

1XFOHDU(QHUJ\,QVWLWXWH 1(, 1(,,QGXVWU\*XLGHOLQHIRU,PSOHPHQWLQJ3HUIRUPDQFH%DVHG2SWLRQRI&)53DUW$SSHQGL[-5HYLVLRQ$GDWHG-XO\Watts Bar-Unit 1 B 3.6-4 Revision 10 Amendment 5

Containment Air Locks B 3.6.2 BASES (continued)APPLICABLE The DBAs that result in a significant release of radioactive material within SAFETY containment are a loss of coolant accident and a rod ejection accident ANALYSES (Ref. 2). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate (La)of 0.25% of containment air weight per day (Ref. 2), at the calculated peak containment pressure of 15.0 psig.This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air Palocks.

 = 15.0 psig The containment air locks satisfy Criterion 3 of the NRC Policy Statement.

LCO Each containment air lock forms part of the containment pressure boundary. As part of containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.(continued)Watts Bar-Unit 1 B 3.6-7 Revision 10 Amendment 5

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.3 (continued)REQUIREMENTS The Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.SR 3.6.3.4 Verifying that the isolation time of each power operated and automatic containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program or in accordance with the Surveillance Frequency Control Program.SR 3.6.3.5 For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 4), is required to ensure OPERABILITY.Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval. Therefore, these valves will be tested in accordance with Regulatory Guide 1.163, which allows a maximum test interval of 30 months. (Ref.3).SR 3.6.3.6 NEI 94-01, Revision 3-A, Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative control.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.(continued)Watts Bar-Unit 1 B 3.6-20 Revision 10, 151, 162 Amendment 5, 123, 132

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.7 REQUIREMENTS (continued) Verifying that each 24 inch containment lower compartment purge valve is blocked to restrict opening to d 50°F is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses of References 1 and 2. If a LOCA occurs, the purge valves must close to maintain containment leakage within the values assumed in the accident analysis. At other times when purge valves are required to be capable of closing (e.g., during movement of irradiated fuel assemblies), pressurization concerns are not present, thus the purge valves can be fully open. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.SR 3.6.3.8 This SR ensures that the combined leakage rate of all Shield Building bypass leakage paths is less than or equal to the specified leakage rate. This provides assurance that the assumptions in the safety analysis are met. The as-left bypass leakage rate prior to the first startup after performing a leakage test, requires calculation using maximum pathway leakage (leakage through the worse of the two isolation valves). If the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange, then the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. At all other times, the leakage rate will be calculated using minimum pathway leakage.The frequency is required by Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria. Although not a part of L a, the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.REFERENCES 1. Watts Bar FSAR, Section 15.0, "Accident Analysis."

2. Watts Bar FSAR, Section 6.2.4.2, "Containment Isolation System Design,"

and Table 6.2.4-1, "Containment Penetrations and Barriers."

3. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, September 1995.

4. Title 10, Code of Federal Regulations, Part 50 Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance - Based Requirements."

Nuclear Energy Institute (NEI) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012 Watts Bar-Unit 1 B 3.6-21 Revision 10, 151, 162 Amendment 5, 123, 132

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential(-2.0 psid) with respect to the Shield Building annulus atmosphere in the event of inadvertent actuation of the Containment Spray System or Air Return Fans.Containment pressure is a process variable that is monitored and controlled.The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.APPLICABLE Containment internal pressure is an initial condition used in the DBA SAFETY ANALYSES analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients. The worst case LOCA generates larger mass and energy release than the worst case SLB.Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. 1).The initial pressure condition used in the containment analysis was 15.0 psia.This resulted in a maximum peak pressure from a LOCA of 9.36 psig. The containment analysis (Ref. 1) shows that the maximum allowable internal containment pressure, Pa (15.0 psig), bounds the calculated results from the limiting LOCA. The maximum containment pressure resulting from the worst case LOCA, does not exceed the containment design pressure, 13.5 psig.(continued)Watts Bar-Unit 1 B 3.6-28 Revision 44, 55, 76, 127 Amendment 33

Enclosure 2 PRA Evaluation CNL-20-006

B45 200306 001 PRA Evaluation Response WBN 0-19-078 Page 1 of 90 IPlant and Unit(s) WBN 1 & 2 Department Requesting Evaluation Licensing Type of Evaluation NOED DCN EOOS > 7 days Unexpected EOOS Results Missed Surveillance Plant Trip(s)Shutdown Evaluation ~ Other WBN CILRT LAR ln~ut Evaluate the risk impact of permanently extending the Containment Type A Integrated Leak Rate Test interval from 1 test-in-10 years to 1 test-in-15 years. The risk impact is characterized as the change in the Large Early Release Frequency (LERF), the increase in the Conditional Containment Failure Probability (CCFP) and the increase in the estimated population dose.Risk Management Actions Required:

None Suggested:
None Conclusion/Discussion/Description of Risk Insights See Attached I'- -

Prepared by Gerry W. Kindred 4 -~ / 6 / 2 0 2 0

 --?' <Print Name/Signature/Date I

 ~r/lY~

Reviewed by Jacob J. Johnson 3/6/2020

 ;;,,fnt Name/Signature/Date TVA 41121 Page 1 of 1 NEDP-26-4 {03-17-2017}

PRA Evaluation Response: WBN-0-19-078 Rev: 000 Plant: WBN Page: 1

Subject:

Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Contents 1.0 Purpose/Background ...................................................................................................4 1.1 References .................................................................................................................................. 5 2.0 Assumptions ................................................................................................................8 3.0 Ground Rules ...............................................................................................................9 4.0 PRA Technical Adequacy for Permanently Extending the Containment CILRT .... 10 4.1 PRA Model Fidelity, Realism and Configuration Control .....................................................10 4.2 PRA Maintenance and Update ................................................................................................11 4.3 PRA Model History ...................................................................................................................11 4.4 Internal Events (With Flooding) PRA Model and Peer Review ............................................14 4.5 Seismic PRA Model and Peer Review ....................................................................................25 4.6 Treatment of Non-Modeled Hazards (Internal Fire) ..............................................................29 4.7 Treatment of Non-Modeled Hazards (High Winds, External Flooding and Other) ............30 4.8 Treatment of Non-Modeled Hazards (Shutdown Events) .....................................................32 4.9 Treatment of FLEX Equipment in the PRA ............................................................................32 4.10 PRA Assessment of Proposed CILRT Interval Extension Methodology ..........................33 4.11 General Conclusion Regarding PRA Capability .................................................................34 4.12 Regulatory Guide 1.174, Revision 3 Defense-In-Depth Evaluation ...................................34 5.0 Methodology...............................................................................................................36 5.1 Step 1 - Baseline Risk Determination ....................................................................................40 5.2 Step 2 - Develop the Baseline Population Dose Per Year ....................................................41 5.3 Step 3 - Evaluate the Risk Impact (Bin Frequency and Population Dose) .........................42 5.4 Step 4 - Evaluate the Change in LERF and CCFP .................................................................43 5.5 Step 5 - Evaluate the Sensitivity of the Results ....................................................................43 6.0 Inputs ..........................................................................................................................44 6.1 Decomposition of LERF Frequency and EPRI Classification ..............................................45 7.0 Calculation .................................................................................................................52 7.1 Step 1 - Baseline Risk Determination ....................................................................................52 7.2 Step 2 - Develop the Baseline Population Dose ...................................................................56 7.3 Step 3 - Risk Impact Evaluation .............................................................................................62 7.4 Step 4 - LERF and CCFP Changes ........................................................................................69 8.0 Sensitivity Analyses ..................................................................................................72 8.1 Liner Corrosion ........................................................................................................................72 8.2 Seismic CDF .............................................................................................................................83 9.0 Evaluation of External Events ...................................................................................83 9.1 Internal Fires Analysis .............................................................................................................83 9.2 Seismic Hazards Analysis .......................................................................................................84 9.3 Other External Events Analysis ..............................................................................................84 10.0 Results/Conclusion..................................................................................................87 10.1 Results Discussion - LERF...................................................................................................87 10.2 Results Discussion - CCFP ..................................................................................................88 10.3 Results Discussion - Population Dose ...............................................................................88 List of Tables Table 1 Internal Events Internal Flooding Model CDF and LERF..................................... 12

PRA Evaluation Response: WBN-0-19-078 Rev: 000 Plant: WBN Page: 2

Subject:

Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 2 IE With IF PRA Model Update Changes ................................................................ 12 Table 3 Seismic PRA Model CDF and LERF...................................................................... 13 Table 4 Seismic PRA Model Update Changes.................................................................. 14 Table 5 IE With IF PRA Model Peer Review SR Capability Category Distribution ......... 14 Table 6 Open IE With IF PRA Open F&Os ........................................................................ 16 Table 7 SPRA Model Peer Review SR Capability Category Distribution ........................ 25 Table 8 Open Seismic PRA Open F&O ............................................................................. 27 Table 9 External Hazards IPEEE and Current Applicability ............................................ 30 Table 10 Detailed Description of EPRI Accident Classes ............................................ 38 Table 11 EPRI Release Classes (Containment Failure Classifications)...................... 45 Table 12 Watts Bar Release Categories and EPRI Mapping ........................................ 45 Table 13 Decomposition of Watts Bar LERF Frequency and EPRI Classification ..... 46 Table 14 Level 2 Accident Sequence Total Frequency....................................................... 52 Table 15 EPRI Accident Class Frequencies ................................................................. 52 Table 16 50-Mile Radius Population Density ................................................................ 57 Table 17 Summary Acciednt Progression Bin (APB) Descriptions ............................ 57 Table 18 Calculation of the Watts Bar Population Dose Risk at 50-Miles .................. 58 Table 19 U1 - Baseline Dose Calculation (Without 3a & 3b) ....................................... 60 Table 20 U2 - Baseline Dose Calculation (Without 3a & 3b) ....................................... 60 Table 21 U1 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution) ... 61 Table 22 U2 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution) ... 61 Table 23 U1 - Testing Once-in-10 Years Risk Profile ................................................... 63 Table 24 U2 - Testing Once-in-10 Years Risk Profile ................................................... 64 Table 25 U1 - Testing Once-in-15 Years Risk Profile ................................................... 64 Table 26 U2 - Testing Once-in-15 Years Risk Profile ................................................... 65 Table 27 U1 - Class 1 PDR Increase Due to Extended Type A CILRT Intervals.......... 66 Table 28 U1 - Class 3a PDR Increase Due to Extended Type A CILRT Intervals........ 66 Table 29 U1 - Class 3b PDR Increase Due to Extended Type A CILRT Intervals ....... 66 Table 30 U1 -Total PDR Increase Due to Extended Type A CILRT Intervals .............. 67

PRA Evaluation Response: WBN-0-19-078 Rev: 000 Plant: WBN Page: 3

Subject:

Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 31 U2 - Class 1 PDR Increase Due to Extended Type A CILRT Intervals ......... 68 Table 32 U2 - Class 3a PDR Increase Due to Extended Type A CILRT Intervals ....... 68 Table 33 U2 - Class 3b PDR Increase Due to Extended Type A CILRT Intervals....... 68 Table 34 U2 Total PDR Increase Due to Extended Type A CILRT Intervals................ 68 Table 35 Unit-1 Summary LERF - CCFP................................................................... 72 Table 36 Unit-2 Summary LERF - CCFP................................................................... 72 Table 37 WBN Liner Corrosion Base-Case Risk Assessment..................................... 75 Table 38 Unit-1 Increase in LERF/yr .............................................................................. 77 Table 39 Unit-2 Increase in LERF/yr .............................................................................. 78 Table 40 Unit-1 Summary of Base Case and Corrosion Sensitivity Cases ................ 81 Table 41 Unit-2 Summary of Base Case and Corrosion Sensitivity Cases ................ 82 Table 42 External Events Contribution to Risk for CILRT Interval Extension ............ 86 Table 43 U1 - Upper Bound on All LERF ....................................................................... 86 Table 44 U2 - Upper Bound on All LERF Contributors ................................................ 87 Table 45 Acceptance Criteria......................................................................................... 87 Table 46 Results Table and Applicability Determination ............................................. 89

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 1.0 Purpose/Background In 1995 the NRC amended 10CFR50 Appendix J to include test methods referred to as Option B, a performance-based approach to containment leakage testing which allows licensees with acceptable test performance history to extend surveillance test intervals. At that time, provisions were made for extending the Containment Integrated Leak Rate Test (CILRT) frequency from the original licensing basis (OLB) of 3 tests-in-10 years to 1 test-in-10 years, supported by the NRCs assessment (NUREG-1493) that stated there is an imperceptible increase in risk associated with the increase in CILRT interval..[5 §10.1.2] During the early 2000s most licenses applied for one-time extensions to 1 test-in-15 years, in accordance with NEI 94-01,[3] including Watts Bar.Integrated leak-rate testing (CILRT) is the only method capable of detecting all existing leaks in the reactor containment system and is only performed during shutdown operations. During the test other activities within or affecting the containment structure cannot be performed; thus, there is an associated cost in terms of critical path, outage duration and lost generation.[5 §1.1]Furthermore, equipment rental to support the test requires several $100Ks, and manpower is also several $100K. Therefore, the overall cost of the test including critical path (~$70,000/hr) is approximately $2,000,000.Industry analysis has shown that, in general, the risk impact associated with CILRT interval extensions for intervals up to 15 years is small; however, plant-specific confirmatory analysis is required. [1, 3 §9.2.3.1] NEI 94-01 Rev 3-A[3] incorporates by reference the EPRI methodology from report 1018243.[1] NRCs Safety Evaluation endorses NEI 94-01 Rev 3-A and is included as part of the upfront material in the NEI guidance document. The purpose of this analysis is to provide the required plant-specific confirmation of permanently extending the Current Licensing Basis (CLB) allowed Containment Type A Integrated Leak Rate Test (CILRT) interval from 1 test-in-10 years to the Proposed Licensing Basis (PLB) of 1 test-in-15 years and represents an insignificant increase in risk The extension would allow for substantial cost savings as the CILRT could be deferred for additional scheduled refueling outages for the Watts Bar plant, which over the remaining life of the plant would reduce the total required number of Type A CILRTs.Earlier assessments followed the guidance of EPRI TR-104285 that considered changes in local leak-rate testing and CILRT testing intervals.[2] That report considered the change in risk based only on population dose, whereas EPRI 1018243 guidance considers population dose, large early release frequency (LERF) and the conditional containment failure probability (CCFP).[1] The risk assessment for Watts Bar follows the guidance from NEI 94-01[3], the methodology used in EPRI TR-1018243,[1] NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in RG 1.200,[11] and RG 1.174[10] for risk insights in support of a request for a change to a plants licensing basis. Additionally, the regulatory endorsed Calvert Cliffs methodology to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval is also used for the Watts Bar analysis.[12]This risk analysis uses the Watts Bar PRA Level 1, Level 2 and Seismic PRA models. The release category and dose (person-rem) information is based on the approach suggested by the EPRI guidance.[1]The NRC report on performance-based leak testing, NUREG-1493[5] analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In the analysis, it was determined that for a representative PWR (i.e., Surry) that containment isolation failures contribute less than 0.1 percent to the latent risk

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval from reactor accidents. This low level of risk contribution is due to the low predicted probability of isolation failure; however, the consequence of containment isolation failure can be substantial.[5,§5.2.2.1]Extending the CILRT interval (Type A test) does not mechanically cause a change in containment isolation leakage, as the Type A test can only identify that there is leakage, if actually present. Furthermore, the increase in the Type A test interval will not cause a significant risk in containment isolation leakage identification because the Type C tests are performed on a more frequent basis than the Type A test.1.1 References

1. EPRI Report 1018243, Oct 2008, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of EPRI 1009325

2. EPRI Report 104285, Risk Impact Assessment of Revised Containment Leak-Rate Testing Intervals

3. NEI 94-01 Rev. 3-A, Industry Guideline for Implementation Performance-Based Option of 10CFR Part 50, Appendix J Reference ML11327A025

4. NUREG-1150, Volumes 1, 2 and 3, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants

5. NUREG-1493, Performance-Based Containment Leak-Test Program

6. NUREG/CR-4551 Vol. 3 Rev. 1, Part 1 Evaluation of Severe Accident Risks: Surry Unit 1

7. NUREG/CR-4551 Vol. 5, Rev. 1, Part 1 Evaluation of Severe Accident Risks:

Sequoyah, Unit 1

8. NUREG/CR-4551 Vol. 5, Rev. 1, Part 2 Evaluation of Severe Accident Risks:

Sequoyah, Unit 1

9. Reg. Guide 1.163 Rev. 0, Performance-Based Containment Leak-Test Program

10. Reg. Guide 1.174 Rev. 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis

11. Reg. Guide 1.200 Rev. 2, An Approach For Determining The Technical Adequacy of Probabilistic Risk Assessment Results For Risk-Informed Activities

12. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C.H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No.50-317, March 27, 2002. ML020920100

13. WBN-0-092, WBN ILRT Level 2 Input Data

14. MDN-000-999-2008-0148 Rev 4, Watts Bar Probabilistic Risk Assessment - Level 2 Analysis

15. ABS Report R-2361441-1823 Watts Bar Unit 2 Severe Accident Mitigation Alternatives

16. Watts Bar Nuclear Plant Individual Plant Evaluation of External Events, April 16, 2015.

17. 1/2-PTI-064-02 , Containment Integrated Leak Rate Test (CILRT)

18. IPEEE, Attachment 4, Internal Fires

19. Watts Bar Nuclear Power Plant, Updated Final Safety Analysis Report, Amendment 113

20. Watts Bar Nuclear Plant Unit-2, Individual Plant Evaluation of External Events, April 16, 2015, Attachment 4, Internal Fires

21. NUREG-1742, Perspectives Gained from the IPEEE Program, USNRC, April 2002

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22. EPRI Report 1025287, November 2012, Seismic Evaluation Guidance

23. ML100270756, GI-199, Appendix D Seismic Core-Damage Frequencies

24. MDN-000-999-2015-000717 Rev 0, WBN Probabilistic Risk Assessment - Seismic Quantification (SQU) Notebook

25. Watts Bar Nuclear Plant Unit 1, Individual Plant Evaluation of External Events, February 1998

26. NEI 05-04 Rev. 3, Process for Performing Internal Events PRA Peer Reviews Using the ASME/AND PRA Standard

27. NPG-SPP-09.11 Rev. 3, Probabilistic Risk Assessment Program

28. NEDP-26 Rev. 12, Probabilistic Risk Assessment

29. MDN-000-999-2008-0151, Rev. 0 Summary Document

30. MDN-000-999-2008-0151, Rev. 1 Summary Document

31. MDN-000-999-2008-0151, Rev. 2 Summary Document

32. MDN-000-999-2008-0151, Rev. 3 Summary Document

33. WBN 1-17-075 Rev. 1 B45 170629 001, Unit 1 Seismic PRA Results

34. MDN-000-999-2015-000717 Rev. 2 Seismic Quantification (SQI) Notebook)

35. WBN-0-18-078, Westinghouse LTR-RAM-II-09-084, RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard requirements for the Watts Bar Nuclear Power Plant Probabilistic Risk Assessment

36. WBN-0-18-079 B45 180614 004, Watts Bar Units 1 & 2 Internal Events Probabilistic Risk Assessment Peer Review Findings Closure

37. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications

38. ASME/ANS RA-Sb-2013 Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications

39. WBN-0-18-080, F&O Closure - Update Versus Upgrade Review

40. NPG-SPP-09.0.4 Rev. 3, Conduct of Probabilistic Risk Assessment Engineering

41. RIMS T42970505066, RIMS Reel E05414, WBN Site Reel 004423915, Problem Evaluation Report WBPER970050, January 22, 1997

42. UFSAR - WBN Updated Final Safety Analysis Report

43. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management

44. CNL-19-035 Response to Request for Additional Information Regarding Application for Technical Specification Change Regarding Risk-Informed Justification for relocating of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

45. CNL-18-067 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

46. NEI 12-13, External Hazards PRA Peer Review Process Guidelines

47. PWROG-16011-P Rev. 0, Peer Review of the Watts Bar Seismic Probabilistic Risk Assessment

48. Jensen Hughes Report 06044-RPT-01, Watts Bar Nuclear Plant Seismic PRA Finding level F&O Independent Technical Review

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49. MDN-000-999-2015-000717 Rev. 0, Seismic Quantification (SQU) Notebook

50. MDN-000-999-2015-000717 Rev. 1, Seismic Quantification (SQU) Notebook

51. ML19210D430 TVA Letter CNL-19-069, Final Response to NRC Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

52. CNL-18-068 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, systems, and Components for Nuclear Power Reactors

53. ML101240992, Watts Bar Nuclear Plant (WBN) Unit 2 - Individual Plant Examination of External Events Design Report

54. CNL-19-035, Response to Request for Additional Information Regarding Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, RAI APLB-05 The following acronyms are used in this calculation:

APB - Accident Progression Bin ARF - Air Return Fans CCDP - Conditional Core Damage Probability CCFP - Conditional Containment Failure Probability CDF - Core Damage Frequency CET - Containment Event Tree CILRT - Containment Integrated Leak-Rate Test CLB - Current Licensing Basis EPRI - Electric Power Research Institute F&O - Facts & Observations HCLPF - High-Confidence of Low-Probability of Failure CILRT - Integrated Leak Rate Test ISLOCA - Interfacing System LOCA La - Leakage (Allowable)LER - Large Early Release LERF - Large, Early Release Frequency MFCR - Mean Fractional Contribution to Risk MOR - Model of Record NEI - Nuclear Energy Institute OLB - Original Licensing Basis PDR - Population Dose-Rate PLB - Proposed Licensing Basis PRA - Probabilistic Risk Assessment RAI - Request for Additional Information RCS - Reactor Coolant System SERF - Small Early Release Frequency

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval SGTR - Steam Generator Tube Rupture STG - Source Term Group WBN - Watts Bar Nuclear Plant 2.0 Assumptions

1. The assumed maximum containment leakage for EPRI Class 1 sequences is 1 La (Type A acceptable leakage) because a new Class 3 has been added to account for increased leakage due to Type A inspections.[1, §5.1.2]

2. The assumed maximum containment leakage (small) for Class 3a sequences is 10 La based on the EPRI guidance. [1, 5.1.2 ]

3. The assumed maximum containment leakage (large) for Class 3b sequences is 100 La based on the EPRI guidance.[1, 5.1.2]

4. Class 3b is conservatively categorized as LERF based on the NEI guidance and previously approved EPRI methodology.[1, §4.2.1.4]

5. Containment leakage due to EPRI Classes 4 and 5 are considered negligible based on the NEI guidance and the previously approved EPRI methodology, and are not evaluated further by this analysis.[1 , Attachment H, §5.1]

6. Conservatively, it is assumed that EPRI Class 8 sequences (ISLOCA, SGTR) are containment bypass sequences; therefore, potential releases are assumed to go directly to the environment.[1 §4.3]

7. A change in the existing 1 test-in-10 years testing frequency to the proposed 1 test-in-15 years frequency assumes a constant failure rate and that the failures are randomly dispersed during the interval between tests. [1 §3.7]

8. It is assumed that a change in CCFP of up to 1.5% is small. This is because NRC has accepted previous submittals with CCFP increase up to 1.1% for one-time extensions of the CILRT testing interval. In context, it is noted that NRC has approved CCFPs of 10%

for evolutionary light water reactor designs.[1, §2.2]

9. The interval between ILRTs at the original licensing basis of 3 tests-in-10 years is taken to be 3 years. This value is consistent with the EPRI guidance report. [1 Table 5-11 Step 3]

10. The likelihood of liner flaw growth over the extended period without benefit of visual inspection is estimated to double every five years. This assumption is generic in nature and does not depend on any plant specific inputs and is used in the EPRI guidance.[1 Table 5-1]

As such, the doubling of the liner flaw likelihood in the Watts Bar analysis is judged to be reasonable.

11. A total visual inspection failure likelihood of 10% is assumed for that fraction of the liner that is available for visual inspection. This assumption is consistent with the EPRI methodology[2 §5.1.5.1] which reads: Five percent failure to identify visual flaws plus 5%

likelihood that the flaw is not visible (not through the cylinder but could be detected by CILRT). All industry events have been detected through visual inspection. Five percent visual failure detection is a conservative assumption.[1 Table 5-11 Note 5]

12. It is assumed that the likelihood of leakage escape due to crack formation in the basemat region is considered to be ten times less likely than the cylinder or dome regions consistent with the EPRI guidance.[1 Table 5-11 Step 4]

13. The containment basemat liner is assumed to be un-inspectable consistent with the EPRI methodology. [1 Table 5-11 Step 5]

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14. Since a larger assumed Containment CILRT pressure yields a worse result in the corrosion sensitivity analysis, an upper bound for containment pressure during an CILRT is used in the analysis. Based on the Containment Integrated Leak Rate test surveillance instruction [17, §5.B.1] the CILRT containment pressure range is 15.0 - 15.6 psig, which corresponds to 29.7 - 30.3 psia. Accordingly, the upper bound pressure selected will be taken to be slightly larger than the 15.6 psig value. This value is considered reasonable because the test range is limited by procedure.[17]

15. Fire events are considered to be the most limiting external hazard due to their frequency of occurrence and their potential impact on plant operation. Therefore, it is assumed that internal fire events bound the risk contribution from other external hazards.

3.0 Ground Rules The following ground rules are used in this analysis:

1. The technical adequacy of the Watts Bar PRA is consistent with the requirements of R.G.

1.200 and is relevant to the Containment CILRT interval extension. The technical adequacy is based on peer review and resolution of the previously open facts &observations (F&Os). All F&Os that did not meet capability category 2 or better have been resolved and transmitted to the NRC in other License Amendment Requests, e.g., TSTF-425.[45]

2. The Watts Bar Level 1 and Level 2 internal events and seismic PRA models provide representative results.

3. It is appropriate to use the Watts Bar PRA models as a gauge to effectively describe the risk change attributable to the ILRT extension. It is reasonable to assume that the impact from the CILRT extension (with respect to percent increases in population dose) will not substantially differ if fire events were to be included in the calculations.

4. Dose results for the containment failures modeled in the PRA can be characterized by information provided in NUREG/CR-4551.[7,8] They are estimated by scaling the NUREG/CR-4551 results by population differences for Watts Bar compared to the NUREG/CR-4551 reference plant.

5. Accident classes describing radionuclide release end-states are defined consistent with the EPRI methodology[1] and are summarized in Section 6.0 of this calculation.

6. The representative containment leakage for Class 1 sequences is 1 La. Class 3 accounts for increased leakage due to Type A inspection failures.[1 §5.1.2]

7. The representative containment leakage for Class 3a sequences is 10 La based on the previously approved methodology for Indian Point Unit 3.[1 §5.1.2]

8. The representative containment leakage for Class 3b sequences is 100 La based on the guidance provided in EPRI Report 1009325 R2.[1 §5.1.2]

9. The Class 3b can be conservatively categorized as LERF based on previously approved methodology.[1 §4.2.1.4]

10. The impact on population doses from containment bypass scenarios is not altered by the proposed CILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, i.e., is not affected by the test interval, no changes on the conclusions from this analysis will result from this separate categorization.

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11. The reduction in CILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal. Therefore, Class 2 impacts remain static.

12. Where possible, this analysis should include a quantitative assessment of the contribution of external events (e.g., fire and seismic) in the risk impact assessment for extended CILRT intervals. For example, where a licensee possesses a quantitative fire analysis and that analysis is of sufficient quality and detail to assess the impact, the methods used to obtain the impact from internal events should be applied for the external event. If the external event analysis is not of sufficient quality or detail to directly apply the methodology provided in NEI 94-01 R3, the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changes interval.

4.0 PRA Technical Adequacy for Permanently Extending the Containment CILRT The Tennessee Valley Authority (TVA) maintains both an Internal Events (IE) PRA Model including Internal Flooding (IF), and a Seismic PRA (SPRA) Model for the Watts Bar Nuclear (WBN) Power Plant. The IE PRA model has been developed in accordance with the requirements of RG 1.200 Rev. 2,[11] subjected to Peer Review[35] and the Appendix X[26] F&O Closure process.[36] The Seismic PRA model (SPRA) was developed and subjected to Peer Review[47] against ASME/PRA Standard RA-Sb-2013,[38 §1.2] which is not endorsed by RG 1.200 Rev. 2, and was subjected to the Appendix X F&O Closure process.[48] See Section 4.5.2 of this analysis for applicability. These models are highly detailed, and include a wide variety of modeled systems, operator actions and common cause events. The TVA PRA uses a multi-faceted, structured approach in establishing and maintaining the technical adequacy and fidelity of the PRA models across its nuclear fleet. This approach includes a proceduralized PRA maintenance and update process [28, §3.2.2], as well as independent peer reviews. The IE with IF PRA quantification process is based on a single top fault tree analysis which is a well-known and accepted methodology in the commercial nuclear power plant industry. The IE with IF model is maintained and quantified using the EPRI Risk & Reliability suite of software programs.The SPRA model is quantified using the widely accepted EPRI FRANX methodology.4.1 PRA Model Fidelity, Realism and Configuration Control WBN PRA model fidelity, realism and configuration control is governed by TVA Fleet procedure NPG-SPP-09.11, Probabilistic Risk Assessment Program[27 §3.2.2] which:

defines PRA model configuration control requirements (e.g., changes to the plant design, operational procedures 1, technical specifications, maintenance and testing, etc.)
defines data collection sources and requirements
defines roles and responsibilities of interfacing organizations (e.g., system engineering, operations, maintenance rule, etc.)

1 Operating procedures include normal operations, emergency operations, off-normal operations, severe accident mitigation guidelines, and others.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 4.2 PRA Maintenance and Update The PRA maintenance and update process is governed by fleet procedures that are applicable to all TVA nuclear units.[28] The TVA risk management process ensures that the applicable PRA models represent the as-built, as-operated plants. Initial models and model upgrades are required to be subjected to independent peer review against the requirements of the ASME/ANS PRA Standard as endorsed by Regulatory Guide 1.200.The following information describes this approach as it applies to the Watts Bar PRA.

NEDP-26, Probabilistic Risk Assessment[28]

o defines the process and management of PRA applications, periodic updates, and model maintenance and review,[28 § 2.0 A]o for risk-informed applications, such as TSTF-425, the procedure requires the PRA staff to revise the appropriate risk related calculations following model updates, [28, §3.1.1 F]o defines PRA Maintenance and PRA Upgrade,[28 §3.2.2 A and B]o updates are required on a routine frequency or by the cumulative impact of plant configuration changes that exceed a threshold value, [28 §3.2.2.D]o the decision to update the PRA model ahead of a normal scheduled PRA maintenance cycle should be made commensurate with the overall impact to the model, taking into consideration the impact on risk-informed applications and programs that use the results from model quantification, for example, Mitigating System Performance Index (MSPI), On-Line Risk Management, SFCP, and others, [28 §3.2.2 E]o defines information sources to review for model updates,[28 §3.2.2 G and H]o defines the living model evaluation of plant changes for the cumulative affect on the PRA results, [28 §3.2.2 J]o PRA model updates are required to be requantified using truncation limits that ensure preservation of model fidelity and to demonstrate convergence for both CDF and LERF. [28 §3.2.3 A]There are two types of updates to the PRA models,

1) PRA Maintenance - the update of PRA models to reflect plant changes such as plant modifications, changes to operating procedures, or plant performance data.

PRA maintenance focuses on ensuring the model accurately reflects the current plant configuration and performance. This includes identification, review, and evaluation of new plant information and the documentation for that information.This is performed at a minimum of once every five years.[28 3.2.2 A]

2) PRA Upgrade - the incorporation into a PRA model of a new methodology or changes in scope of capability that significantly impacts the results of the significant accident progression sequences. [28 §3.2.2 B]

4.3 PRA Model History The original Watts Bar IE and IF PRA was developed using the RISKMAN platform. In 2009 the PRA model was completely rebuilt using CAFTA.[29 §3.0] All documentation for the Internal Events and Internal Flooding models were upgraded to meet the requirements of RG 1.200 Rev. 2.[29] The model history shown in section 4.3 begins with

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval CAFTA model rev. 0, which was subjected to peer review (discussed in section 4.4.1).Section 4.3.2, Table 2 IE With IF PRA Model Update Changes provides the major changes associated with each model revision.4.3.1 Internal Events PRA with Internal Flooding Table 1 Internal Events Internal Flooding Model CDF and LERF U2 U1 Model Date U1 CDF/yr U2 LERF/yr CDF/yr LERF/yr Rev. 0[29] NOV 2010 1.82E-05 1.71E-05 1.64E-06 1.65E-06 Rev. 1[30] FEB 2014 1.39E-05 1.43E-05 1.12E-06 1.16E-06 Rev. 2[31] SEP 2016 1.06E-05 1.03E-05 1.44E-06 1.43E-06 Rev. 3[32] MAR 2017 9.08E-06 8.98E-06 1.01E-06 1.00E-06 4.3.2 Internal Events PRA With Internal Flooding - Model Updates Table 2 IE With IF PRA Model Update Changes

 % Change Model Change Comments In LERF In CDF [Table 2]

[Table 2]Changes made in the CAFTA Rev. 0 model include:[ 30 §6.1.1.3]

added components to fault tree modeling to represent the as-built plant

 -23.5 (U1) -31.7 (U1)

model refinements of component Rev. 1[30] configurations to represent the as-operated

 -26.4 (U2) -29.7 (U2) plant

revised the HRA dependency approach to retain individual HEPs (human error probabilities) in cutsets
internal flooding flood frequencies were updated to the EPRI-TR-1021086 values Changes made to the CAFTA Rev. 1 model include: [Level 1 31 §6.1.14.1] [Level 2 31 §6.1.14.2]
model changes to reflect component boundaries described in the Data Analysis Notebook
Common Cause Factors (CCFs) updated from Multiple Greek Letter (MGL) to the Alpha

 -23.7 (U1) 28.6 (U1) Factor Methodology Rev. 2[31] -28.0 (U2) 23.3 (U2)

Test basic events and Maintenance basic events have been merged to a single basic event, Test & Maintenance (TM)
the component cooling system (CCS) model and the support system initiating event (SSIE) model were revised
FLEX diesel generators were added to the model

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 % Change Model Change Comments In LERF In CDF [Table 2]

[Table 2]

MAAP runs were added or revised to represent the as-built, as-operated plant
hydrogen detonation probabilities were updated
the probabilities of small and large pre-existing containment leaks were updated
the probabilities of pressure-induced and temperature-induced steam generator tube ruptures were updated
Level 2 binning was updated to reflect changes to the Plant Damage States (PDS) for each CDF sequence defined in the Accident Sequence Notebook Changes made to the CAFTA Rev. 2 model include:

 -14.3 (U1) -29.9 (U1)

AC power modeling was enhanced to Rev. 3[32]

 -12.8 (U2) -30.1 (U2) represent the as-built plant to crosstie power across the units in the event of a loss of offsite power 4.3.3 Pending Model Updates Affecting IE With IF PRA The PRA pending model change database for potential model changes has been reviewed and no potential change meets the criteria for a non-scheduled PRA model update, and no pending changes meet the criteria for a model upgrade.

4.3.4 Seismic PRA Table 3 Seismic PRA Model CDF and LERF Model Date U1 CDF/yr U2 CDF/yr U1 LERF/yr U2 LERF/yr Rev. 0[49] FEB 2016 N/A 4.31E-06 N/A 2.21E-06 Rev. 1[50] ] APR 2017 N/A 2.69E-06 N/A 1.53E-06 May 2017 (U2)Rev. 2 2.60E-06[33] 2.61E-06[34] 1.70E-06[33] 1.73E-06[34]Jun 2017 (U1)

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 4.3.5 Seismic PRA - Model Updates Table 4 Seismic PRA Model Update Changes Model % Change % Change Comments In CDF In LERF N/A (U1) N/A (U1)

Revision 1 reflects changes made in response Rev. 1[50] to resolving F&Os

 -37.6% (U2) -30.8% (U2)

N/A (U1) N/A (U1)

Revision 2 reflects changes made as part of the Rev. 2[33, 34] closure review process

 -3.0% (U2) +13.1% (U2) 4.4 Internal Events (With Flooding) PRA Model and Peer Review 4.4.1 Internal Events With Internal Flooding PRA Peer Review Assessment The Watts Bar Internal Events with Internal Flooding PRA was subjected to a full scope peer review in November 2009,[35 §1.3] in accordance with the requirements of NEI 05-04[26] Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard. [35 §1.1] The review covered all technical elements from the ASME/ANS PRA Standard Parts 2 and 3 plus the configuration control element. [37 §3.0]

Table 5 IE With IF PRA Model Peer Review SR Capability Category Distribution Capability Number[45 §2.4.3 Table 5] Per-Cent*Category Not Met 26 8.0 I 203 62.3 II 19 5.8 III 34 10.4 I/II 5 1.5 II/III 14 4.3 I/II/III 16 4.9 Not Applicable 9 2.8 TOTAL: 326 100%

 *Rounded The conclusion of the peer review team is as follows:[35 §5.0]

The overall model structure is robust and well-developed, but needs refinement,
Documentation is very thorough, detailed and well organized such that comparison with the standard is facilitated,
The processes and tools utilized for the WBN PRA are at the state of technology and generally consistent with Capability Category II,

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The PRA maintenance and update program includes all necessary processes and does a very good job of tracking changes, and
The qualitative assessment of sources of modeling uncertainty for the Level 1 model is very comprehensive and well documented to support future applications.

The peer review was performed against the RA-Sa-2009 standard which was endorsed by RG 1.200 Rev. 2.[11 App A] The Peer Review Team stated that the Watts Bar PRA meets the ASME/ANS PRA Standard. The PRA has issues which have been documented with Facts and Observations (F&Os).[35 §5.0] A total of 50 finding level F&Os were prepared in this review.4.4.2 Internal Events With Internal Flooding PRA F&O Closure Results The 50 finding level F&Os identified by the 2009 peer review team were subjected to a Peer Review Closure process in June 2017 over a four day period at the TVA Chattanooga offices. The review was performed in accordance with the process documented in NEI 05-04 Appendix X, as well as the requirements published in the ASME/ANS PRA Standard (RA-Sa-2009) and Regulatory Guide 1.200 Revision 2.[36 §1.1]A team of three independent PRA experts performed the F&O reviews along with consensus sessions consisting of the entire team.[36] The review met the Appendix X requirement that each F&O review include two qualified reviewers. Furthermore, the team examined the changes made to Watts Bar PRA models, data, and documentation to address the findings to determine if the Capability Category II (or better) requirements of the ANS/ASME PRA Standard, including clarifications imposed by Regulatory Guide 1.200, Revision 2 are met.[36 §1.1]The Closure Peer Review Team had significant PRA experience, and each team member confirmed they were not TVA employees, had no involvement in development of the WBN PRA or performance of risk applications for WBN, and no conflicts of interests, incentives or disincentives.[36 Tables B-1, B-2]The F&O closure peer review team concluded that 43 F&Os met the requirements for closure, leaving seven open F&Os, which are discussed in Table 6 of this evaluation. The team determined that none of the PRA updates made to address the F&Os were determined to be PRA upgrades and no new PRA methods were utilized. The closed F&Os fell into the following general categories: [36 §3.0]

Instances in which minor errors in the fault tree modeling were noted, which were corrected,
Instances in which changes to common cause basic events were needed to address specific issues,
Instances in which the PRA documentation did not adequately discuss the details of the PRA models and data, hence, resolution involved adding further documentation details,
Instances in which the method for modeling human error dependency was changed; however, this change primarily affected only how the recovery rules were applied to adjust for dependencies. Numerically, the results are identical to the previous method; only the process used to assign the dependency adjustments was changed,

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Instances in which specific data or modeling details were missing or incorrect in certain parts of the model, but were determined to be appropriate in other parts of the model,
Instances in which documentation of sources of modeling uncertainty were not included. While the documentation of these sources of uncertainty are important (particularly for applications), the addition of this documentation (which was reviewed by the finding closure review team) is needed for completeness and does not impact the PRA model itself or the risk results,
Instances where some of the uncertainty parameters for specific basic events were not included. However, the majority of the events had uncertainty parameters and the uncertainty analysis itself was satisfactorily performed.

TVA evaluated each closed F&O against the endorsed ASME/ANS PRA Standard examples for upgrade versus maintenance criteria.[37 Appendix 1-A.3] This review confirmed the peer review closure teams conclusion that no actions taken to close the F&Os met the criteria of an upgrade.[39]Table 6 provides the F&Os for the Internal Events Model with Internal Flooding that remain open along with the potential impact on the CILRT extension.Table 6 Open IE With IF PRA Open F&Os F&O 1-6 MDN-000-999-2008-0145 [Data Analysis] Section 5.3 documents the Bayesian update process used for WBN. Both mean and EF values are produced for each type code. However, it was noted that uncertainty interval data was not entered into the WSBN2.RR file and that extraneous information from previous versions of the database were being applied to the factor (demands or exposure time) field of the BE table.Associated BASIS FOR SIGNIFICANCE (SRs) Incorrect entry of uncertainty intervals in the CAFTA database will result DA-D3 in incorrect output from the UNCERT program.POSSIBLE RESOLUTION Review the WSBN2.RR file to ensure appropriate uncertainty interval information is entered for each type code and that the uncertainty interval information in the basic event table is removed where it is not applicable.PLANT RESPONSE Uncertainty data was added into the *.rr file of the WBN model.Uncertainty error factors confirmed removal from basic event table where not applicable.Closure STATUS Review OPEN BASIS While most of the entries in the .TC and .RR files contain uncertainty information, there still appear to be some entries [that] miss such information. Uncertainty information is not included in the .RR file for

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval dependent HEP combinations, recovery sequences (e.g., L2FXSBO01 and XSBO01), failure of RHR legs (e.g., RHRCLA-PB), failure of SI piping (e.g., SICL-PB), common cause combinations generated by the CCF tool (e.g., U0_032_ACAS_CMP_FR_CCF_1_2 and U2_EPS_GA_27_FD_CCF_1_2), and ISLOCA basic events (e.g.,U2_62-662RV). Uncertainty information is not included in the TC file for operator actions, HXCPL_CC4A3, SLOCA, and SSBO.Impact on Quantification of the PRA model uses point estimates for the the CILRT calculation of CDF and LERF; therefore, uncertainty information as Extension described in the closure review would not cause a change to those results. Although it is recognized that some omissions of uncertainty information for some entries is present, they would not be used in the PRA analyses to support the CILRT extension.F&O 2-28 MDN-000-999-2008-0144 Appendix F addresses identification of dependencies. The criteria are met since the analysts followed common practice. However, the stated rule for application of a lower limit (1E-05) on the combined HEP was not applied in the Qrecover File.Associated BASIS FOR SIGNIFICANCE (SRs)Some of the combined operator action probabilities are below the HR-D5 threshold specified in the notebook.HR-G7 POSSIBLE RESOLUTION QU-C1 Redefine the lower threshold for combined HEPs to a value of 1.0E-06 and ensure the combined HEP values are consistent with this threshold.QU-C2 The basis for the lower limit could be that some of the PSFs are global in nature and apply as a sum rather than a product. For any combinations which are retained with a value lower than the specified threshold, a justification should be provided.PLANT RESPONSE Appendix F of MDN-000-999-2008-0144 was moved to the Quantification Notebook (MDN-000-999-2008-0147). Combined HEPs were limited to 1E-5 in Appendix F of MDN-000-999-2008-0147 revision 5.Closure STATUS Review OPEN BASIS Appendix F states: "In order to satisfy this, the recovery rule file was developed to limit the joint probability of each combination to be no less than 1.0E-05." However, review of dependency recovery file U1_CDF_HRA_DEP_MOR_R3.RECV revealed that not all HEP combinations were replaced with a probability no less than 1E-05. A few examples of combinations with probabilities less than 1.0E-05 still exist:COMBINATION_256_U1C 5.499E-07, COMBINATION_257_U1C

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 4.195E-08, etc. These same values are reflected in the summary table of Appendix F.Impact on The CILRT is not affected by HEPs as the calculation determines the the CILRT change in LERF, Dose, and CCFP. Additionally, use of the JHEP floor Extension values was shown in a sensitivity study (WBN-1-19-075) to result in no change in the CDF or LERF cutsets. Therefore, the impact on the CILRT application is negligible; however, TVA intends to review the recovery rule file and address the JHEPs lower than the acceptable floor value.F&O 3-6 Section 5.8 of the Quantification Notebook provides a result of the parametric uncertainty analysis. The analysis does not include the uncertainty parameters for the CCF events and ISLOCA events. In addition, the HRADEP* recovery events found in the recovery files are not treated properly in the parametric uncertainty analysis.Associated BASIS FOR SIGNIFICANCE (SRs)The parametric uncertainty assessment is only a partial assessment.QU-A3 The assessment needs to properly account for the CCF events, ISLOCA events and HRA events in the parametric uncertainty assessment, or QU-E3 provide a State-Of-Knowledge Correlation assessment to show that the results are not impacted significantly.POSSIBLE RESOLUTION Either include the CCF events, ISLOCA events and HRA events properly in the parametric uncertainty assessment, or provide a State-Of-Knowledge correlation assessment to show that the results are not impacted significantly. The concern with uncertainty assessment of the CCF events is that uncertainty parameters are not defined for the MGL factors. Therefore, the uncertainty analysis only propagates the uncertainty parameters of the independent failures to the CCF events.Consideration should be given to adopting the Alpha method (which does allow definition of uncertainty parameters for each factor) or performance of additional sensitivity analysis to assess the correlated uncertainty of the CCF events.PLANT RESPONSE The uncertainty analysis is documented in MDN-000-999-2009-0162.ISLOCA uncertainty is discussed in section 5.4.2.6, CCF uncertainty is discussed in Section 5.8, and HRA uncertainty is discussed in section 5.7.Closure STATUS Review OPEN

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval BASIS Discussion of sources of modeling uncertainty for ISLOCA, CCF, and HEP dependent events are included in the current version of the Uncertainty and Sensitivity notebook (MDN-000-999-2009-0162).However, this F&O and the associated SRs pertain to parametric uncertainty. The Uncertainty and Sensitivity notebook specifically notes that the state of knowledge correlation was not applied in the ISLOCA calculation of valve failure probabilities (as required by QUA3). It does not appear that the items identified in this F&O have been addressed.Therefore, this F&O remains open.Impact on This circ*mstances of this F&O does not affect the CILRT evaluation the CILRT as the parameters in question do not affect the delta risk calculated in Extension support of the extension.F&O 5-8 The operator action failure probabilities considered in the LERF analysis were not correctly estimated. After core damage, the operation steps in the SAMGs would be much different from the steps in the EOPs before core damage.Associated BASIS FOR SIGNIFICANCE (SRs)HAPRZ is a key operator action to prevent high pressure accident LE-C2 scenarios. HAPRZ was estimated to be 4.4E-04 while a similar operator action for the level 1 analysis, HAOB1, was estimated to be 1.6E-02.LE-C7 POSSIBLE RESOLUTION LE-C9 Describe more specifically how the HEP for action HAPRZ was LE-E1 calculated and how the calculation accounted for conditions after core damage.PLANT RESPONSE No changes were made. The F&O comments particularly focused on event HAPRZ and its HEP relative to event HAOB1. The action HAPRZ is an in-control room action that assumes that the execution stress is high as does action HAOB1. The control room conditions would not be different post-core damage versus pre-core damage (lighting, heat, humidity, radiation and atmosphere). These are not expected to change in EOIs versus SAMG scenarios. The actions associated with HAPRZ are not complex, consisting of opening all pressurizer PORVs and block valves. As to the comparison between HAOB1 and HAPRZ, the system time window (Tsw) and the time available for recovery is shorter for HAOB1 (30 minutes and 12.5 minutes, respectively) than for HAPRZ (1.4 hours and 73 minutes, respectively). The execution steps to perform HAPRZ are not as involved as those required to perform HAOB1 leading to a smaller execution error probability for HAPRZ.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Status OPEN Closure Review BASIS Based on actions taken, SR LE-C9 may be considered to be MET at capability category II-III; SR LE-C2 remains met at Cat I because there are several examples of Operator Actions following the onset of core damage that were treated conservatively and not updated to address the F&O (e.g., operator actions to recover equipment were not specifically considered, actions to blind feed the steam generators were not credited, and manual actuation of the nitrogen backup system were not credited). As a result, F&O 5-8 cannot be considered closed.This finding addressed over-conservatisms with respect to operator Impact on actions that are not credited following the onset of core damage. The the CILRT CILRT extension risk metrics focus on the delta risk of an additional Extension five year window of containment corrosion, therefore, the absence of the non-credited operator actions is not expected to have an appreciable impact on the delta LERF or population dose. It would be expected to have no impact on the conditional containment failure probability.F&O 7-10 The analysis in Section 5.4.1 [Internal Flooding Notebook] includes an assessment that evaluates existing human actions. From a cursory review, the main impact seems to be an exclusion of non-MCR actions given a flood event. There appears to be little if any adjustment to the other actions that are performed in the MCR.Associated BASIS FOR SIGNIFICANCE (SRs) The information in Table 5-15 lists the existing operator actions and IFQU-A6 defines an impact. No changes are listed for MCR events and those not in the MCR are typically considered to be infeasible. The text indicates that "All actions solely performed from the Main Control Room (MCR) are also expected not to be physically impacted by the flood event." This seems to be in contrast to the SR requirement to adjust PSFs to address additional stress and the work environment following a flood event. This is particularly of interest for events that could include damaged systems such as starting a CCP (HACV2) which could increase flooding rates or results in failure of standby equipment.POSSIBLE RESOLUTION Develop a more detailed assessment of why no change would be anticipated for actions or perform a PSF evaluation concentrating on those events that could compound the event (fail equipment due to lack of cooling for instance).PLANT RESPONSE MCR actions were considered; generally these were determined not to require changes. Factors such as stress, cues, effect of flood on

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval response timing etc., were considered and discussed in Section 5.4.1.Scenario specific information was also considered. The detailed assessment provides information why no change is required for some actions, and why other actions are changed. (see Table 5-12 and 5-13).STATUS OPEN BASIS Confirmed that the discussion in Section 5.4.1 provides the information identified in the F&O response, however, no impacts of these actions were able to be identified. It is noted that the justifications for MCR actions all identify EPRI Guidance items 3.4 & 4.3 as the basis for making no changes to MCR HEPs; however, both items 3.4 & 4.3 state they only apply to HFEs for which there is more than one hour available for the operator action. The MCR actions listed in the table vary from 10 min. to 14.4 hr. For actions that have greater than 1 hour for response or are only Closure in response to IEs that cannot be caused by a flood initiator, items 3.4 &Review 4.3 are valid justification for no change to the HEP. For the HEPs that are required in less than 1 hour and required to respond to a flood initiator, items 3.4 & 4.3 are not valid and, in these cases, items 4.1 & 4.2 should apply and these require consideration of an increased stress level to account for flood stress levels and consideration of an increased median response time to account for the flood effects, e.g., higher stress and flood distractions. Based on the current documentation, it is not possible to determine whether/how these latter effects were considered to reach the conclusion that there are no numerical effects on the MCR HEPs that are required for a flood initiators in less than 1 hour. As a result, the actions implied by the F&O resolution response cannot be confirmed for these HEPs and the F&O remains open.Impact on This finding addresses a documentation issue that does not support the CILRT review in a manner to determine, or to confirm the conclusion of no Extension numerical effects. The CILRT extension analysis is not affected by the issue discussed in this finding because the analysis determines the change in risk for a longer test interval. In addition, it was shown that there is a negligible impact of a flood event on those HFEs which would require less than an hour to either diagnose or perform is presented in the response to RAI APLA-04 in CNL-19-035.F&O 7-21 The range factors are developed for the flood initiating events, however there is no propagation through the model.Associated BASIS FOR SIGNIFICANCE (SRs)The current analysis does include uncertainty estimates for the flood IFEV-B3 initiating events. However, the impact and resultant uncertainty associated with combining the different flooding sources, each with an associated range factor, with regard to the overall study uncertainty is not addressed. Additionally, the sensitivity of assumptions related to propagation and flow rates with regard to consequential failures should

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval be addressed to ensure that the impact of such simplifications on the overall results are known.Associated BASIS FOR SIGNIFICANCE (SRs)The current analysis does include uncertainty estimates for the flood IFEV-B3 initiating events. However, the impact and resultant uncertainty associated with combining the different flooding sources, each with an associated range factor, with regard to the overall study uncertainty is not addressed. Additionally, the sensitivity of assumptions related to propagation and flow rates with regard to consequential failures should be addressed to ensure that the impact of such simplifications on the overall results are known.POSSIBLE RESOLUTION Perform a statistical uncertainty assessment for the results and provide additional sensitivity studies assuming various combinations of assumptions related to initiating event grouping and consequences.PLANT RESPONSE The current internal flooding notebook includes the parametric uncertainty values developed for the initiating events. The CAFTA basic event database also includes uncertainty parameters for the flooding initiators.STATUS OPEN BASIS The flooding initiating basic events contained in the current CAFTA .rr file include the error factors and distribution information. The values shown are consistent with the error factors noted in Table 5-16 of the internal flooding notebook. However, it appears that pipes of multiple sizes (and possibly piping from various systems) were lumped together into a single initiating event. It is not clear from the documentation how Closure the selected error factor was calculated in cases where different error Review factors are shown for various pipe sizes.SR IFQU-A7 states: "PERFORM internal flood-induced accident sequence quantification in accordance with the requirements described in 2-2.7 as applicable to flood-induced accident sequences." Section 2-2.7 dictates how quantification is to be performed for internal events and includes SR QU-E3 which requires parametric uncertainty be performed. While the WBN IF logic was included in the parametric uncertainty analysis using UNCERT, the absence of error factors for many of the basic events (see F&O 3-6) tends to nullify the validity of the analysis.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Impact on This finding addresses a documentation issue that does not support the CILRT review of how the selected error factor was calculated in certain cases.Extension The CILRT analysis extending the test interval is not affected by the issue discussed in this finding because the CILRT evaluation determines the change in risk for a longer window of vulnerability. In addition, the response to RAI APLA-05 in CNL-19-035 demonstrates that the impact of the missing error factors does not result in a significant change to the results.F&O 7-22 The secondary side isolation of a ruptured SG was modeled in the SGTR event tree (top event SL). After core damage, there was no consideration of the secondary side isolation capability in the accident progression sequences.Associated BASIS FOR SIGNIFICANCE (SRs)A cycling SRV allows for the SG to be maintained at a higher pressure LE-D5 which tends to increase holdup time prior to release to the environment and to reduce the rate of release such that the overall source term is lower than for cases with a stuck open SG SRV on the faulted steam generator. Prior analyses have indicated that the resulting reduction is sufficient to reduce the source term from large to small.Associated (SRs)LE-D5 POSSIBLE RESOLUTION The analysis of the SGTR sequences should include credit not only for the ability to maintain covered tubes, but also the impact of the SG SRV cycling instead of failing open. This would provide a sizeable reduction in the release and may result in the reclassification of some LERF sequences to SERF.PLANT RESPONSE The SR LE-C4 was considered met Cat II for this element. This action is considered to be an enhancement. Current sequences which may not result in LERF may currently be counted as LERF, possibly resulting in conservative results. Assumption 30 from rev. 5 of the level 2 notebook addresses F&O 7-20 and F&O 7-22: "After core damage, there is no consideration of the secondary side isolation capability in the accident progression sequences.A cycling SRV allows for the SG to be maintained at a higher pressure which tends to increase holdup time prior to release to the environment and to reduce the Level 2 analysis assumes that all core damage sequences that have feedwater available will result in a small early release. However, a review of the core damage cutsets indicates that the dominant SGTR sequences are due to failure of long term heat removal, which would actually probably all be late releases. The accident binning conservatively bins all SGTR sequences to either small early or large early releases, possibly resulting in conservative results."

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Furthermore, as noted in the response to RAI-08.b.i[51 the intent of this F&O is to question why secondary side isolation was not credited in the LERF analysis. In order to assess the potential for masking SSC risk ranking, a sensitivity study was performed that applied a 0.1 recovery factor to the applicable SGTR sequences to evaluate the impact to LERF. It was determined that this sensitivity study resulted in no change to the LERF cutsets, meaning the importance measures were not affected.Closure STATUS Review OPEN BASIS Confirmed that Level 2 NB includes an assumption that there is no consideration of secondary side isolation capability, even though it is recognized that there is likely source term reduction if SG PORVs /MSSVs are cycling vs stuck-open. Addition of this assumption has no impact on the level of realism used to model secondary side isolation for SGTR sequences, so SR LE-D5 remains MET at Cat I and the F&O resolution is considered to be not closed.This finding addresses a conservatism with respect to not crediting Impact on secondary side isolation capability. The CILRT analysis for extending the CILRT the CILRT interval is not affected by the issue discussed in this finding Extension because the evaluation determines the change in risk for the longer window of vulnerability of containment liner corrosion. The change in risk associated with this finding is shown in the response to RAI 8b.i in CNL-19-069 is shown to have a negligible impact to the PRA results.Therefore there is no impact of modeling the aforementioned secondary containment isolation capability.Closure Review Documentation[36 App A]F&O Documentation[35 App. C.1]4.4.3 Pending Model Updates Affecting IE With IF Modeling TVA Fleet procedure, Conduct of Probabilistic Risk Assessment Engineering[40]prescribes the process to ensure the PRA models represent the as-built, as-operated plant configurations in support of integrated decisionmaking and maintaining a high sensitivity for reactor safety in all activities, actions and responses. The requirement for both permanent and temporary changes to the plant design or operation is assessed [40 §3.2.1] by PRA Engineers that monitor changes to the plant design and operating procedures for impact on the PRA model. The PRA Engineer is responsible for ensuring that design changes that impact the PRA model are appropriately incorporated into the PRA model.[40

 §3.2.9]

Changes in PRA inputs or discovery of new information are required to be evaluated to determine whether such information warrants PRA update (including the cumulative effect of all previously evaluated model changes that are yet to be included in the MOR).Changes causing CDF or LERF to exceed +/- 25% requires an off-cycle model update.[28

 §3.2.2]

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 4.5 Seismic PRA Model and Peer Review 4.5.1 Seismic PRA Model and Peer Review The 2016 Watts Bar Seismic PRA (SPRA) peer review was performed in accordance with NEI 12-13 External Hazards PRA Peer Review Process Guidelines,[46 §1.1] The scope of the peer review included all technical elements in Part 5 [Requirements for Seismic Events At-Power PRA] of the ASME/ANS RA-Sb-2013 PRA Standard, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. See 4.5.2 for discussion on the acceptability of using PRA Standard Addendum B.Part 5 of Addendum b of the PRA standard is referenced in the Electric Power Research Institute (EPRI) report, Screening, Periodization and Implementations Details (SPID) for the Resolution of f*ckushima Near-Term Task Force Recommendation 2.1: Seismic which the NRC has endorsed as one acceptable method for responding to the information requested in Enclosure 1 of the 50.54(f) letter pertaining to Post-f*ckushima Near Term Task Force (NTTF) Recommendation 2.1 on seismic re-evaluation.[22 §1.2]A diverse team of eight experts converged at the TVA offices the week of March 14, 2016.In accordance with the requirement of independence (section 1-6 2.2 of the ASME/ANS PRA Standard), each expert certified they did not participate in the development or participation of any portion of the Watts Bar SPRA that they reviewed. Furthermore, each expert confirmed the Peer Review report reflects the process used by the team, and the element grading, facts, observations and conclusions as agreed to by the review team in its consensus discussions during the review.[48 §1.1] The distribution of F&Os with respect to Capability Category is presented below.Table 7 SPRA Model Peer Review SR Capability Category Distribution[47 Table 4-1]Capability Number[48 Table 4-1] Per-Cent*Category Not Met 10 12%2 (U1) 2% (U1)I 1 (U2) 1% (U2) 18 (U1) 21% (U1)II to CC-III 19 (U2) 22% (U2)I/II/III 51 59%Not Applicable 5 6%TOTAL: 86 100%

 *Rounded

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval As noted in letter CNL-19-035[44] For the WBN TSTF-425 submittal RAI responses, the staffs May 1, 2017 letter providing comments on the F&O Independent Assessment is addressed in Table 1 of the response to RAI ALPB-01.Furthermore, TVA letter to NRC, CNL-19-035[44] With TVAs RAI ALPB-01 response, on the NRC staff letter from March 7, 2018 on NEI guidance 12-13 with respect to comments on external hazards PRA peer Review process are addressed in Table 1 of the response to RAI ALPB-01.The conclusion of the peer review was that, in general, the data, methodologies and seismic risk models used for WBN Units 1 & 2 were appropriate and sufficient to meet the majority of the ASME/ANS PRA Standard requirements. All but ten supporting requirements were met. A total of 74 findings were identified by the review team. Based on this peer review, the team concluded the SPRA is judged to be essentially consistent with the PRA Standard and can be used for risk-informed applications. Furthermore, the team stated that the technical adequacy of the WBN SPRA is very good.[48 §5]4.5.2 Acceptability of Peer Review Against the 2013 Non-Endorsed PRA Standard RG 1.200, Revision 2 endorses ASME/ANS RA-Sa-2009 (Addendum A) but, as noted in an NRC letter (ML111720067) to ASME, NRC does not endorse PRA Standard ASME/ANS RA-Sb-2013 (Addendum B). As noted in Section 2.5.1 of the WBN License Amendment Request (LAR) to adopt TSTF-425,[45] the SPRA model was subjected to a peer review against the Part 5 (Seismic) supporting requirements of the non-endorsed ASME/ANS RA-Sb-2013 PRA Standard. The SPRA peer review was performed in accordance with NEI 12-13, External Hazards PRA Peer Review Process Guidelines.No exceptions to use of NEI 12-13 were noted in the peer review report.As noted in TVA letter CNL-19-035 to NRC which provides the responses to the TSTF-425 PRA application RAIs, the differences between PRA Standard Addendum A and Addendum B are discussed in terms of how the supporting requirements in Addendum B are consistent with the endorsed standard for this application. Where the different criteria are not consistent with the endorsed standard, discussion is provided on how the analogous Addendum A supporting requirements have been met. This information is documented in Table 1 of the response to RAI APLB-02, TVA letter CNL-19-035.[44]Table 1 provides supplemental information to the Vogtle letter for Addendum A supporting requirements not addressed by Addendum B.4.5.3 Seismic PRA F&O Closure Review The WBN SPRA Finding Level F&O Technical Review was performed at the Jensen-Hughes offices from April 10 - 13, 2017. The purpose of the review was to perform an independent assessment to review TVAs close out of Finding level F&Os of record from the WBN SPRA peer review against the ASME/ANS PRA Standard, Addenda B.The technical review team consisted of six team members and a dedicated team lead, all of which have extensive qualifications and many years of experience in the pertinent areas of SPRA and peer review. The review met the Appendix X[46] requirement that each F&O review include two qualified reviewers.[48 §2.2.2] All reviewers met the criteria

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval specified NEI 12-13 and the ASME/ANS RA-SA-2009 PRA Standard Section 1-6.2, including independence. [48 §Executive Summary]The review process was based on the previously completed SPRA peer review. The closure review is intended to support WBN License Amendment Requests (LAR) submittals and other regulatory interactions. Finding level F&O dispositions reviewed and determined to have been adequately addressed through the closure review are considered closed; and no longer relevant to the current PRA model, and, therefore, do not need to be carried forward nor discussed in the CILRT LAR submittal.[48 §Executive Summary]The F&O closure team reviewed the 74 finding level F&Os from the SPRA Peer Review and determined that all but one were judged to be resolved, and therefore, closed. The one remaining finding level F&O was technically partially resolved and is discussed in section 4.5.4 for applicability to the proposed risk-informed CILRT application. [48 §Executive Summary]4.5.4 Applicability of Seismic Open Peer Review Findings (F&O)According to the Peer Review Closure Team only one F&O, SHA 20-5 remains open.[48

 §4.0]

The F&O is characterized as Technically Resolved; however, remains open due to documentation. [48, Appendix B]Table 8 Open Seismic PRA Open F&O F&O 20-5 Analyses have been performed that indicate that in numerous scenarios[48, Appendix B] involving the seismic-induced failure of earthen and concrete dams upstream of WBN, the resulting flood evaluation at WBN does not exceed 728 ft. However, these analyses are not adequately summarized in the SPRA documentation.Associated BASIS FOR SIGNIFICANCE (SRs) The document entitled "Position Paper on Other Seismic Hazards, Watts Bar Nuclear Plant Unit 2" describes the potential for flooding at WBN due to seismic-induced dam failures upstream of the plant and cites the SHA-I1 FSAR as the source of the supporting analyses. The analyses in the FSAR are obsolete and have been replaced by a more comprehensive set of analyses based on current information for seismic hazard, dam stability, and flood routing (Calculation DQ0000002014000024). These newer analyses indicate that in all but the most conservative (and unlikely) scenario, the resulting flood will be less than 728 ft (i.e., plant grade). For the one scenario that yields a flood evaluation above 728 ft, the plant has approximately 30 hours to initiate the appropriate response.POSSIBLE RESOLUTION A summary of the recent analyses regarding seismic-induced dam failure should be incorporated into the seismic hazard documentation.PLANT RESPONSE A summary of the recent analyses regarding seismic-induced dam failure has been incorporated into CDN0000002015000739, Revision 1, Appendix III. [Seismic Induced Dam Failure and Flooding]

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Closure STATUS Review Technically Resolved - Open Documentation BASIS A revised version of Appendix III [Seismic Induced Dam Failure and Flooding] summarizes the methodology and results of the analyses used to evaluate the potential for flooding at WBN due to the potential seismic failure of upstream dams. The analyses summarized in Appendix III indicate that one scenario (involving the failure of multiple dams upstream of WBN as a result of ground motions equal to 1/2 of 10-4 AFE coincident with a 500-yr flood) results in a flood that exceeds plant grade (728 ft). For this scenario, Appendix III notes that several conservative assumptions were made regarding the timing of the failure of the immediate upstream dam (Watts Bar) and the assumed stability of the downstream dam (Chickamauga Dam). Appendix III also points out that flood protection for the plant is provided to elevation 738.9 feet and that procedures are in place to respond to a flood warning within 27 hours.For the scenario that exceeds plant grade, over 30 hours warning is available. Finally, Appendix III also provides results that indicate if more realistic, less conservative assumptions were made for the scenario that exceeds plant grade, the resulting flood elevation is reduced by nearly 7 feet to elevation 723 feet (5 feet below plant grade).The preponderance of evidence provided in Appendix III and in supporting documentation supports a conclusion that failure of upstream dams during a seismic event is unlikely to lead to flooding that could not be mitigated at the plant and thus can be screened out as a potential seismic hazard. However, the case that is made does not systematically recognize, discuss and address the potential sources of uncertainty in the dam breach process, in the estimation of flooding levels and thus does not present a complete case for screening out this hazard based on the results of the analyses.SUGGESTION

1. State clearly the criterion that will be used to screen out seismic induced dam failures.

2. Identify the elevation that will be used to define the plant flood capacity; discuss what plant grade is and what the flood protection level of the plant is. Also discuss what you will use as the basis for the screening analysis and why.

3. Identify the governing dam failure events (combinations) that will be considered in the analysis and the basis for their selection.

4. Systematically identify and describe the sources of aleatory and epistemic uncertainty in the analysis, with particular emphasis on those that will impact the estimate of peak flood elevations at the plant.

5. Ideally, a realistic/best estimate analysis of dam failures would be carried out. Note, the analysis that was presented does not seem

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval (though the reviewers are not 100% sure) to be a realistic best estimate analysis. A best estimate would have realistic consideration of the breach characteristics of a dam, realistic timing of dam failures, etc.

6. Identify and evaluate how different sources of aleatory uncertainty could impact the results (possibly estimating the size of these uncertainties).

7. Identify and evaluate how different sources of epistemic uncertainty could impact the results and how these are addressed.

Impact on The F&O Closure Peer Review Team concluded that the issue the CILRT associated with F&O 20-5 has been technically resolved and pertains Extension only to a suggestion for documentation enhancements. As such, there is no change that the proposed suggestions to the SPRA CDF or LERF.With no change to CDF or LERF, there is no impact on the PRA assessment of evaluating the impact of the CILRT extension.4.5.5 Pending Model Updates Affecting Seismic PRA The PRA pending model change database for potential model changes has been reviewed and no potential change meets the criteria for a non-schedule SPRA model update, and no pending changes meet the criteria for a model upgrade.4.6 Treatment of Non-Modeled Hazards (Internal Fire) 4.6.1 Internal Fires Unit 1 Fire Induced Vulnerability Evaluation (FIVE)WBN performed a Fire Induced Vulnerability Evaluation (FIVE) in accordance with EPRI TR-100370. This methodology consisted of a progressive screening approach based on quantifying the following:[16 §1.3]

Fire ignition frequency for specific plant areas
Availability of automatic suppression systems
Availability of redundant/alternate safe shutdown systems
Zone of Influence (ZOI) of ignition sources
Detailed evaluation of safe shutdown components and their availability The screening was conducted in 3 phases:
Fire Area Screening (Qualitative Analysis)
Critical fire Compartment Screening (Quantitative Analysis)
Plant Walkdown / Verification and Documentation The Internal Fire analysis did not uncover any serious fire vulnerabilities and no plant modifications were recommended as a result of the study. In keeping with the requirements of Supplement 4 to Generic Letter 88-20 (NUREG-1407) and the guidance provided by the EPRI FIVE (TR-100370) documentation, this evaluation

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval confirmed that there were no fire-induced vulnerabilities associated with the continued operation of Watts Bar Nuclear plant.[16 §1.4.2]Unit 2 Fire Induced Vulnerability Evaluation (FIVE)The Unit-2 assessment of fire hazards is an extension of, and used a methodology consistent with, the evaluation performed for Unit-1.[20 §4]4.7 Treatment of Non-Modeled Hazards (High Winds, External Flooding and Other)WBN performed the screening described in Supplement 4 to Generic Letter 88-20, and NUREG-1407. Because WBN was designed prior to the 1975 Standard Review Plan (SRP), the approach included review of the design bases and compared them to the 1975 SRP requirements. [16, 25 §1.3] The IPEEE was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.The IPEEE evaluation revealed that the plant meets the 1975 SRP criteria for other external hazards and only one recommendation was issued by the team.[25 §1.4.3] The issue was documented and dispositioned by the Corrective Action Program (CAP).[41]In addition to extreme winds and tornadoes, external flooding including intense local precipitation and transportation and nearby industrial facilities were evaluated,[25 §8.3] and documented in Attachment 5 of the Unit 1 IPEEE. Selection of external events for the IPEEE and the technical approach recommended for evaluation of such external events are discussed in Section 2 and Section 5 of NUREG-1477, respectively. The recommendation was based on those external initiators that have the potential of initiating an accident that may lead to severe reactor core damage or large radioactive release to the environment.The external events required to be evaluated by NRC in the IPEEE response include:[25 Att 5]

Seismic (replaced by the Seismic PRA)
Internal Fires
High Winds and Tornadoes
External Floods
Transportation and Nearby Facility Accidents A review was performed of the external events described in NUREG-1407 and other external events further required each licensee to confirm that no plant-unique external events are excluded from the WBN IPEEE. [25 §5.2]

Table 9 External Hazards IPEEE and Current Applicability Event WBN IPEEE Applicability Current Applicability A modified site specific Replaced by the Seismic PRA Seismic program was used in the Model IPEEE. [25 Table 3-1]

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The conclusions if the IPEEE FIVE analysis remain valid.Subsequent to the FIVE The EPRI Fire Induced analysis additional modifications Vulnerability Evaluation (FIVE) have been made to the plant approach was used in the WBN including fire wrap, cable tray IPEEE. It was confirmed that covers, refined fire response Internal Fires there are no fire-induced procedures, etc.; therefore, the vulnerabilities associated with as-built as-operated plant is continued operation of WBN[25 considered more robust with Att 4 §6]respect to impacts from internal fires as compared to the plant evaluated in the IPEEE FIVE analysis.The basis for screening High CAT I structures have been Winds and Tornadoes, remains designed to resist tornado wind applicable to the current plant and missile effects equivalent design. 0-AOI-8 provides to the 1975 SRP criteria. SSCs instruction for actions to be High Winds, important to safety were taken in the event of a Tornado Tornadoes designed to withstand the Watch or a Tornado Warning design basis tornado and issued for Rhea or Meigs remain functional. There are no County. These actions further unique vulnerabilities for high mitigate the potential winds. [25 Att 5 §5.5.3] consequences of a tornado event.External flooding due to Design meets the NRC upstream dam failures and regulatory position 2 of RG maximum potential precipitation 1.59. The new PMP (Probable has been re-evaluated since the Maximum Precipitation) criteria analysis performed for the External was evaluated and WBN was IPEEE. Furthermore, Floods designed to withstand this flood enhancements have been made and prevents water from to the upstream dam and other entering safety-related plant modifications to limit the structures.[25 Att 5 §5.6.4] impact on the plant due to external floods.There are no significant industrial facilities near WBN.The impact of potential The nearest land transportation Transportation transportation and nearby is State Route 68 about one-mile and Nearby facility accidents were north of the plant. The nearest Facility evaluated and concluded that main railroad line is about seven Accidents their contribution to plant risk is miles from the plant. No other negligible. [25 Att 5 §5.7.3] significant industrial land use, military facilities, or transportation routes are in the

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval vicinity of WBN.[USFAR §2.2.1] An evaluation of potential accidents concluded that no activities are being performed in the vicinity of the plant that could be considered hazardous to the plant. A study of products and materials transported past the site by rail or barge indicates that no potential explosion hazard exists. Postulated accidents resulting in fire or dense smoke will no effect plant safety.[42 §2.2.3]All external other external hazards have been screened based on generic data (e.g.,Other external hazards are Lightning), site location (e.g.,Other External judged to remain screened Volcanic Activity), generic Hazards based on the same criteria used bases (e.g., Extraterrestrial for screening by the IPEEE.Activity), low probability of occurrence (e.g., Turbine Missiles). [25 Table 3-1]4.8 Treatment of Non-Modeled Hazards (Shutdown Events)Shutdown events are not applicable to the CILRT interval extension as events postulated under shutdown operations are not affected by the CILRT, regardless of frequency.Shutdown events do not affect the metrics associated with the CILRT interval extension.4.9 Treatment of FLEX Equipment in the PRA 4.9.1 Methodology to Assess Failure Probabilities of FLEX Equipment[54]The Internal events and seismic PRA models credit the permanently installed FLEX Emergency Diesel Generators (EDGs). Use of these EDGs was modeled similarly to other internal event PRA components, including development of fragilities, data, and operator actions.WBN does not include portable FLEX equipment in the PRA models. WBN includes the permanently installed FLEX Diesel Generators within the PRA model and supporting components including fuel tank, alignment of breakers, buses, and operator actions to align the FLEX Diesel Generators. The failure probabilities for the equipment was assumed to be the same as other components of the same type already included within the model (e.g.the FLEX diesels have the same failure probabilities as the Emergency Diesel Generators).The NRC memorandum dated May 30, 2017, "Assessment of the Nuclear Energy Institute 16-06, 'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269),

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval is generally focused on portable FLEX equipment. In regard to data analysis it states the following:ASME/ANS PRA Standard Capability Category II (CC-II) for supporting requirement (SR) DA-D1 states that realistic parameter estimates for significant basic events should be calculated based on relevant generic and plant-specific evidence unless it is justified that there are adequate plant-specific data to characterize the parameter value and its uncertainty. Parameter estimates for the remaining events should be calculated by using generic industry data. Supporting Requirement DA-D2 states that if neither plant-specific data nor generic parameter estimates are available for the parameter associated with a specific basic event, USE data or estimates for the most similar equipment available, adjusting if necessary to account for differences.Alternatively, USE expert judgment and document the rationale behind the choice of parameter values.The WBN internal events and seismic PRA models have followed the guidance of the ASME/ANS PRA standard in crediting the permanently installed FLEX DGs. In addition, the FLEX DGs are an uncertainty identified in the PRA model and are subjected to a sensitivity analysis when applicable for the STI extension under consideration.4.9.2 HRA Approach for FLEX Strategies[54]WBN used the HRA Calculator to address the human performance shaping factors for each of the FLEX Human Error Probabilities (HEPs). Each type of Diesel Generator has a separate HEP, as the method to start and align the diesels are different. Both of these HEPs were developed in accordance with the corresponding technical elements in the NRC endorsed ASME/ANS PRA Standard.WBN has procedures governing the initiation into mitigating strategies for use of the permanently installed FLEX diesel generators. The procedures are explicit in what steps must be performed for these actions which are modeled in the PRA in a similar fashion as other human failure events (HFEs)..4.10 PRA Assessment of Proposed CILRT Interval Extension Methodology The TVA process for analyzing the proposed extension of the CILRT interval follows the endorsed NEI 94-01 guideline.[3]4.10.1 Key PRA Attributes The evaluation allows for a blended approach to assessing the change in risk associated with a change to the CILRT interval extension. This includes approved probabilistic risk models as well as non-PRA methodologies (e.g., FIVE).Hazards to be evaluated

Internal Events at Full Power (which includes internal flooding)
Internal Fire Events
Seismic Events
Other External Hazards (e.g., high winds, flooding, etc.)
Shutdown Events

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 4.10.2 I 4.11 General Conclusion Regarding PRA Capability The Watts Bar PRA maintenance and update process and technical capability evaluations (Peer Reviews) described above provide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions, including the CILRT interval extension.4.12 Regulatory Guide 1.174, Revision 3 Defense-In-Depth Evaluation Regulatory Guide 1.174, Revision 3 (Reference 12), describes an approach that is acceptable for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. One of the considerations included in Regulatory Guide 1.174, Revision 3, is defense-in-depth. Defense-in-depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility.The seven considerations presented in Regulatory Guide 1.174, Revision 3, Section 2 1 1 2, "Considerations for Evaluating the Impact of the Proposed Licensing Basis Change on Defense in Depth," are used to evaluate the proposed licensing basis change for overall impact on defense-in-depth. Each of the seven considerations are presented below with a PRA response provided.Current Technical Specification 5 5 15 indicates that for Type C tests the containment leakage rate testing program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A NEI 94-01, Revision 3-A, guidelines allow extension of the containment Type C test interval up to 75 months, based on acceptable performance.The impact of Type C testing in accordance with NEI 94-01, Revision 3-A, was considered in the following defense-in-depth evaluation.

1. Preserve a reasonable balance among the layers of defense A reasonable balance of the layers of defense (that is, minimizing challenges to the plant, preventing any events from progressing to core damage, containing the radioactive source term, and emergency preparedness) helps to ensure an apportionment of the plant's capabilities between limiting disturbances to the plant and mitigating their consequences. The term "reasonable balance" is not meant to imply an equal apportionment of capabilities. The NRC recognizes that aspects of a plant's design or operation might cause one or more of the layers of defense to be adversely affected. For these situations, the balance between the other layers of defense becomes especially important when evaluating the impact of the proposed licensing basis change and its effect on defense-in-depth.

PRA Response

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The usage of the risk metrics of large early release frequency (LERF), population dose, and conditional containment failure probability (CCFP) collectively ensures the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved The change in large early release frequency is small per Regulatory Guide 1 17 4, and the change in population dose and conditional containment failure probability are small as defined in Attachment B of this submittal and consistent with NEI 94-01, Revision 3-A.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures PRA Response The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small change in risk associated with the Type A test frequency change.

3 Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.PRA Response The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change 4 Preserve adequate defense against potential common-cause failures PRA Response Adequate defense against common-cause failures is preserved. The Type A test detects problems in the containment, which may or may not be the result of a common-cause failure. Such a common-cause failure may affect failure of another portion of containment (that is, local penetrations) due to the same phenomena. Adequate defense against common-cause failures is preserved via the continued performance of the Type B and Type C tests and the performance of inspections. The change to the Type A test, which bounds the risk associated with containment failure modes including those involving common-cause failures, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.

5. Maintain multiple fission product barriers PRA Response Multiple Fission Product barriers are maintained. The portion of the containment affected by the Type A test extension is still maintained as an independent fission product barrier, albeit with a small change in the reliability of the barrier. Fuel cladding and the reactor coolant system (RCS) are not affected by the proposed change.

6 Preserve sufficient defense against human errors

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval PRA Response Sufficient defense against human errors is preserved. The probability of a human error to operate the plant, or to respond to off-normal conditions and accidents is not significantly affected by the change to the Type A testing frequency. Errors committed during test and maintenance may be reduced by the less frequent performance of the Type A test (less opportunity for errors to occur).7 Continue to meet the intent of the plant's design criteria PRA Response The intent of the plant's design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or the way the plant is operated. As the intent of the plant's design criteria will continue to be met, this consideration for evaluating the impact of the proposed change on defense in depth will not affect the risk associated with the proposed change.Conclusion The responses to the seven defense-in-depth questions above conclude that the existing defense-in-depth has not been diminished. Therefore, the proposed change does not comprise a reduction in safety.5.0 Methodology The methodology employed is in accordance with NEI 94-01, Revision 3-A[3] and the NRC regulatory guidance on the use of PRA and risk insights in support of a license amendment request (LAR) for changes to a plants licensing basis, R.G. 1.174[10] This methodology is similar to that presented in the EPRI guidance[1] as specified in NEI 94-01.[3]A simplified bounding analysis approach is used in the methodology to evaluate the risk impact on increasing the CILRT Type A interval from the current licensing basis of 1 test-in-10 years to the proposed licensing basis of 1 test-in-15 years by examining specific accident sequences in which the containment remains intact or those in which it is impaired. The aspects considered included:

Accident progression sequences in which the containment remains intact initially and in the long term (Class 1)

 - Class 1 Frequency 2 = FREQINTACT - FREQClass 3a - FREQClass 3b where; Class 3a = small containment liner leakage Class 3b = large containment liner leakage 2 The adjustment to Class 1 is necessary to maintain the sum of the frequencies equal to CDF.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval

Accident progression sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B 3 or Type C 4 tested components. For example, steam generator manway cover leakage.
Accident progression sequences in which containment integrity is impaired due to containment isolation failures due to pathways (e.g., misalignment) left open following a plant post-maintenance test.
Accident progression sequences 5 involving containment failure by any of the following:

 - Large Containment Isolation Failures (Class 2) - Small Containment Isolation Failure-to-Seal Events (Class 4 and 5) - Containment Isolation Failures - Dependent Failures, Personnel Errors (Class 6) - Severe Accident Phenomena Induced Failures (Class 7) - Containment Bypass Events (Class 8)

Section 5 contains the following tables and sections.Table 10 presents detailed information regarding the EPRI accident classes[1, §4.3]Step 5.1 discussion on how the baseline risk is determined Step 5.2 discussion on how the baseline population dose/yr is determined Step 5.3 discussion on how the risk impact (Bin Frequency & Population Dose) is determined Step 5.4 discussion on the how the change in LERF and CCFP is determined Step 5.5 discussion on how the sensitivity is determined including CCFP and External Events 3 Type B tests measure component leakage across pressure retaining boundaries, e.g., gaskets, expansion bellows and air locks.4 Type C tests measure component leakage rates across containment isolation valves.5 The sequences of these classes are impacted by changes in Type B and Type C test intervals, and are not affected by changes in the Type A test interval.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 10 Detailed Description of EPRI Accident Classes EPRI Population Dose Population Dose-Rate Description [1, §4.3]Frequency Leakage Class (person-rem) (person-rem/rx-yr)CONTAINMENT INTACT - all core damage accident progression bins for which the containment remains intact with negligible leakage. Class 1 sequences arise from those core Calculated Value damage sequences where containment isolation is successful Value from 1 and long-term containment heat removal capability is available. La NUREG/CR- DOSEClass 1

FClass 1 FClass 1 =

The frequency of an intact containment is established on the 4551 CDFIntact - F3a - F3b individual plants PRA. For Class 1 sequences, it is assumed that the intact containment end-state is subject to a containment leakage rate less than the containment allowable leakage (La).LARGE CONTAINMENT ISOLATION FAILURES - all core damage accident progression bins for which a pre-existing From Plant PRA Value from From Plant 2 leakage due to failure to isolate the containment occurs. These FClass 2 = NUREG/CR- DOSEClass 2

FClass 2 PRA sequences are dominated by failure-to-close of large (>2 PLargeCI
CDFTotal 4551 diameter) containment isolation valves.

SMALL PRE-EXISTING LEAK IN CONTAINMENT - all core Calculated Value damage accident progression bins with a pre-existing leakage FClass 3a = (Class 1 dose for 3a in the containment structure in excess of normal leakage. PClass 3a

CDF 10 La DOSE3a
F3a La)
10 Small leaks are characterized as > 1 La 10 La.

LARGE PRE-EXISTING LEAK IN CONTAINMENT - all core Calculated Value damage accident progression bins with a pre-existing leakage FClass 3b = (Class 1 dose for 3b 100 La DOSE3b

F3b in the containment structure in excess of normal leakage. PClass 3b* CDF La)
100 Large leaks are characterized as > 10 La.

SMALL ISOLATION FAILURE - FAILURE TO SEAL (TYPE B TEST) - all core damage accident progression bins for which a failure-to-seal containment isolation of Type B test components occurs. Because these failures are detected by Type B tests 4 N/A N/A N/A N/A and their frequency is very low compared with the other classes, this group is not evaluated further. The frequency of Class 4 sequences is subsumed into Class 7, where it contributes insignificantly.SMALL ISOLATION FAILURE - FAILURE TO SEAL (TYPE C TEST) - all core damage accident progression bins for which a failure-to-seal containment isolation of Type C test components 5 N/A N/A N/A N/A occurs. Because these failures are detected by Type C tests and their frequency is very low compared with the other classes, this group is not evaluated further. The frequency of

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval EPRI Population Dose Population Dose-Rate Description[1, §4.3] Frequency Leakage Class (person-rem) (person-rem/rx-yr)Class 5 sequences is subsumed into Class 7, where it contributes insignificantly.CONTAINMENT ISOLATION FAILURES (DEPENDENT FAILURES AND PERSONNEL ERRORS) - similar to Class 2.These sequences involve core damage accident progression bins for which failure-to-seal containment leakage, due to failure 6 N/A N/A N/A N/A to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following test/maintenance evolutions. i.e., human error. All other failure modes are bounded by the Class 2 assumption.SEVERE ACCIDENT PHENOMENA - INDUCED FAILURES - From Plant PRA all core damage accident progression bins in which FClass 7 = Value from From Plant 7 containment failure induced by severe accident phenomena CDFCFL + CDFCFE NUREG/CR- DOSE7

F7 PRA occurs (e.g., hydrogen combustion and direct containment 4551 heating).

CONTAINMENT BYPASS - all core damage accident From Plant PRA progression bins in which containment bypass occurs. Each FClass 8 = Value from From Plant 8 plants PRA is used to determine the containment bypass CDFISLOCA+ NUREG/CR- DOSE8

F8 PRA contribution. Contributors include ISLOCA and SGTR CDFSGTR 4551 (unisolated) events.

CDFIntact = core damage frequency for intact containment sequences from the plant-specific PRA PLarge CI = random containment large isolation failure probability (i.e., large valves)CDFTotal = total plant-specific core damage frequency PClass 3a = the probability of a small (10 La) pre-existing containment leak PClass 3b = the probability of a large (100 La) pre-existing containment leak CDFCFL = the core damage frequency resulting from accident sequences that lead to late containment failure CDFCFE = the core damage frequency resulting from accident sequences that lead to early containment failure

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The risk metrics used to evaluate the impact of a proposed change on plant risk include the following figures of merit and acceptance criteria:The figures of merit (or risk metrics) [1, page 2-4]:

the change in LERF (LERF),
change in risk as defined by the changes in dose (Population Dose [Person-Rem]),
total LERF,
and, the change in the conditional containment failure probability (CCFP).

The acceptance criteria:

LERF and Total LERF, RG 1.174[10]
Population Dose<1.0 person-rem or <0.1% Increase - whichever one is less restrictive[1, §1.2]
CCFP 1.5% (Assumption 8)[1, §1.2]

The Type A containment test measures the ability of the containment to maintain its function, therefore, the proposed change has no measureable effect on the Level 1 PRA core damage frequency (CDF). The Level 1 PRA CDF remains constant and has no risk significance with respect to the containment CILRT test interval.The overall methodology[1 , §4.2] used in this analysis followed these steps:

1. Define and quantify the Baseline Risk Determination

2. Develop the Baseline Population Dose

3. Evaluate the Risk Impact (Bin Frequency and Population Dose)

4. Evaluate Change in LERF and CCFP

5. Evaluate Sensitivity of Results 5.1 Step 1 - Baseline Risk Determination This step[1, §4.2.1] is to define and quantify the baseline risk in terms of core damage frequency (CDF) for each EPRI accident class, excluding classes 4, 5 and 6. According to the EPRI guidance these accident classes are excluded because the circ*mstances (i.e., CILRT Type B and Type C tests) and types of failures such as simultaneous failure of redundant isolation valves are not impacted by changes in the CILRT Type A frequency.[1] The baseline risk is determined as follows:

The plant-specific Watts Bar Level 2 PRA release categories [13] are mapped to EPRI accident classes 2, 7 and 8. This is accomplished by linking the release category definitions to the appropriate EPRI accident class.
The release categories that represent accident Class 1, Containment Intact, are those identified as not having containment failure. The Watts Bar Level 2 analysis refers to these as INTACT. The release categories representing INTACT, or no containment failure outcomes may experience leakage due to the increased window of vulnerability of extending the test interval. The increase in leakage contribution is subtracted to obtain the expected no containment failure outcome frequency as follows:

FREQClass 1 = CDFINTACT - FREQClass 3a - FREQ Class 3b

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval To adjust the Class 1 frequency it is necessary to maintain the sum of the frequencies of the accident classes equal to the total CDF.

Class 3 end-states are developed specifically for this application. These end-states include all core damage accident progression bins with a pre-existing leakage in the containment structure in excess of normal leakage.[1, §4.3] The frequencies for Class 3a and Class 3b are determined as follows:

FREQClass 3a = CDF

Class 3a leakage probability (PClass_3a)

FREQ Class 3b = CDF

Class 3b leakage probability (PClass_3b)

Class 3a represents containment liner leakage characterized as small. The probability is based on industry data. Class 3b represents containment liner leakage that is large which has a probability based on Jeffreys Non-Informative Prior.[1, §3.5]According to the EPRI guidance[1] The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and thus not associated with the postulated Type A containment large leakage path (LERF). The contributors can be removed from class 3b in the evaluation of LERF by multiplying the class 3b probability by only that portion of CDF that may be impacted by Type A leakage.

An example of the type of sequences that may independently cause LERF is a sequence associated with containment bypass events, such as steam generator tube rupture (SGTR) or interfacing system loss of coolant accidents (ISLOCA). Another example may include those accident sequences associated with anticipated transients without SCRAM (ATWS) events.
An example of the type of sequences that may never result in LERF is a sequence where containment sprays and containment heat removal are available. In these sequences, containment sprays and cooling reduce the fission products via scrubbing and rapidly reduce containment pressure. The basis for the removal of sequences to reduce conservatism is plant and PRA specific and should be documented by analysis in the risk impact assessment.[1, §4.2.1]

Core damage accident progression end-states are developed for the Watts Bar PRA Level 2 results.[13] which are used to define the representative sequences. Based on the discussion above, determination of the Type A CDF contribution involves identifying two different scenarios.

1) those scenarios corresponding to release categories which include unmitigated containment bypass or pre-existing large isolation failures and 2) those release categories where there is no containment isolation failures prior to core damage combined with effective mitigation of fission product releases. There is no containment isolation failure prior to core damage.

5.2 Step 2 - Develop the Baseline Population Dose Per Year In step 2[1, §4.2.2] the baseline dose/yr corresponding to the current licensing basis CILRT testing interval (1 test-in-10 years) is estimated. Watts Bar specific estimates of population dose were

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval developed in support of Severe Accident Mitigation Alternatives (SAMAs) for completion and operation of WBN Unit 2 based on 2000 Census data.SECPOP2000 uses data from the 2000 Census to provide population estimations for input to winMACCS. The population estimates provided for the Severe Accident Mitigation Alternatives calculation were projected for the year 2040 using growth rates from the county population projections. [15, §4.5.2]The yearly population dose is estimated for each accident class by multiplying the dose estimate for a class by either the frequency estimated in Step 1 or the La factor corresponding to the Class.[1,§4.2.2]

1. From the Watts Bar specific Level 2 results,[13] determine the relationship between offsite dose measured in person-rem and containment leakage rate (the dose in person-rem) for Class 1. Assumed to be equal to 1 La.

2. From the plant Individual Plant Examination of External Events (IPEEE),[16] determine the offsite dose (person-rem) for the accident classes where analysis is available, typically Classes 1, 2, 7 and 8.

3. For those accident classes where analysis is not available in the IPEEE or PRA, determine the dose estimate by determining the class containment leak rate and multiplying by the 1.0 La dose.

4. The offsite dose estimate for EPRI accident Classes 3a and 3b are estimated as following in accordance with the EPRI guidance.

3a = Class 1 (1 La)

10 3b = Class 1 (1 La)
100

5. Determine the baseline accident class dose-rates (person-rem/yr) by multiplying the dose by the frequency for each of the accident classes. Sum the accident class dose-rates to obtain the total dose-rate.

5.3 Step 3 - Evaluate the Risk Impact (Bin Frequency and Population Dose)In this step,[1, §4.2.3] the risk impact associated with the change in CILRT intervals is described.

1. Determine the change in probability of leakage detectable only by CILRT (Classes 3a and 3b) for the new surveillance intervals of interest. NUREG 1493[5] states that relaxing the CILRT frequency from 3 tests-in-10 years to one in ten years will increase the average time a leak goes undetected by an CILRT from 18 to 60 months (1/2 the surveillance interval), 60/18 = 3.33 fold increase. Therefore, relaxing the CILRT testing frequency from 3-tests-in-10 years to 1-test-in-15 years will increase the average time a leak goes undetected by an CILRT from 18 to 90 months (1/2 the surveillance interval), 90/18 = 5.0 fold increase.

2. Determine the population dose-rate for the new surveillance intervals of interest by multiplying the dose by the frequency for each of the accident classes. Sum the accident class dose-rates to obtain the total dose-rate.

3. Determine the increase in dose-rate and percentile increase for each extended interval as follows: Increase in dose-rate = (total dose-rate of new interval minus total baseline dose),

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval and percent increase = [(increase in dose-rate) divided by (total baseline dose-rate)] x 100%.5.4 Step 4 - Evaluate the Change in LERF and CCFP In this step,[1, 4.2.4] the changes in LERF and CCFP are described.Section 5.4.1 Change in LERF is described.Section 5.4.2 Change in CCFP is described.5.4.1 Evaluate Risk Impact - Change in LERF The risk associated with extending the CILRT interval involves a potential that a core damage event that normally would result in only a small radioactive release from containment could result in a large release due to an undetected leak path enlarging during the extended interval. Only Class 3 sequences have the potential to result in early releases if a pre-existing leak were present.Late releases are excluded regardless of the size of the leak because late releases are not, by definition, LERF events. The frequency of class 3b sequences is used as a measure of LERF, and the change in LERF is determined by the change in class 3b frequency. Refer to Regulatory Guide 1.174[10] for LERF acceptance guidelines.LERF = (frequency class 3b interval x) - (frequency class 3b baseline) 5.4.2 Evaluate Risk Impact - Change in CCFP Evaluate the change in CCFP. The conditional containment failure probability is defined as the probability of containment failure given the occurrence of a core damage accident, which can be expressed as:CCFP = [1 - (frequency that results in no containment failure)/CDF]

100%

CCFP = [1 - (frequency class 1 + frequency class 3a)/CDF]

100%

CCFP Change (increase) = (CCFP at interval x) - (CCFP at baseline interval),expressed as percentage point change.5.5 Step 5 - Evaluate the Sensitivity of the Results In this step,[1, §4.2.5] the sensitivity of the risk impact results to assumptions in liner corrosion are investigated.

Evaluate the sensitivity of the impact of extended intervals to liner corrosion. The methodology developed for Calvert Cliffs[12] investigates how an age-related degradation mechanism can be factored into the risk impact associated with longer CILRT testing intervals. The instances of through-wall penetration flaws are considered in the development of the risk assessment methodology and are part of the plant-specific analyses performed for assessing the potential for liner corrosion.
As stated in the Calvert Cliffs analysis,[12] occurrences of through wall liner corrosion related defects had been found between September 1996 implementation of the visual inspection requirements of 10CFR50.55a and the submittal date for that reference. The defects were found in the cylinder region of the liner. No defects were identified in the basemat region.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 5.5.1 Containment Overpressure The Watts Bar plant does not rely on containment overpressure to aid in net-positive suction head (NPSH) for emergency core cooling system (ECCS) injection.5.5.2 External Events Where possible, the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended CILRT intervals. For example, where a licensee possesses a quantitative fire analysis and that analysis is of sufficient quality and detail to assess the impact, the methods used to obtain the impact from internal events should be applied for the external event. If the external event analysis is not of sufficient quality or detail to allow direct application of the methodology provided in this document, the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses (e.g., FIVE) or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval.The EPRI guidance[1,§5] provides an example of the technical approach for the assessment of external events LERF. Watts Bar will use the Seismic PRA for the seismic hazard, and the FIVE evaluation performed for the IPEEE for internal fires.6.0 Inputs In this section inputs from the Watts Bar Level 2 PRA are provided and the relationship to the corresponding EPRI accident class is given.Table 11 presents the EPRI release classifications and the interpretation for assignment to the Watts Bar release categories.Table 12 presents the Watts Bar release categories, descriptions and mapping to the corresponding EPRI accident class.Section 6.1 presents the decomposition of the Watts Bar accident sequences and EPRI classification.Table 13 presents the decomposition of the Watts Bar accident sequences, frequencies and corresponding EPRI classification.Table 15 presents the EPRI accident class frequencies, the Total CDF and the Baseline CDF (which excludes Class 6).To determine how the Watts Bar release categories relate to the eight EPRI accident classifications the definitions are interpreted and documented in Table 11.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 11 EPRI Release Classes (Containment Failure Classifications)EPRI Interpretation for Assigning Description[1 Table 4-1]Class Watts Bar Release Category Containment remains intact with containment 1 Intact accident sequence bins initially isolated Isolation faults that are related to a loss of power 2 Large containment isolation failures or other isolation failure mode that is not a direct failure of an isolation component Independent containment isolation failures due to Isolation failures identified by Type A testing, 3Type A related failures Large (3b) or Small (3a)Independent containment isolation failures due to 4 Isolation failures identified by Type B testing Type B related failures Independent containment isolation failures due to 5 Isolation failures identified by Type C testing Type C related failures Other penetration failures (dependent failures or Isolation failure with scrubbing or small isolation 6personal errors) failures Early and Late containment failure sequences as 7 Induced by severe accident phenomena a result of hydrogen detonation or other early phenomena 8 Bypass Bypass sequences, ISLOCA or SGTR The Watts Bar Level 2 accident sequences are parsed into seven release categories that represent the summation of individual accident categories due to similar characteristics. Table 12 presents the seven release categories, descriptions[14] and mapping to the corresponding EPRI class.[1, §4.3]Table 12 Watts Bar Release Categories and EPRI Mapping Release EPRI Description Category Class BLERF Large Early Release (LER) Via Bypass of Containment 8 HLERF LER - High Pressure Sequences 7 ILERF LER - Containment Isolation Failures 2 INTACT Containment Intact - No Release 1 LATE Late Release - All Scenarios 7 LLERF LER - Low Pressure Sequence 7 SERF Small Early Release 6 6.1 Decomposition of LERF Frequency and EPRI Classification The decomposition of the Watts Bar accident release categories into the individual Level 2 accident sequences,[13] corresponding EPRI class and the frequency is provided in Table 13.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 13 Decomposition of Watts Bar LERF Frequency and EPRI Classification L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class BLERF-001 3.82E-09 3.94E-09 8 BLERF-002 2.33E-10 2.72E-10 8 BLERF-003 3.96E-08 3.96E-08 8 BLERF-004 4.53E-07 4.46E-07 8 BLERF-005 1.80E-07 1.78E-07 8 HLERF-001 3.13E-08 3.29E-08 7 HLERF-002 <5.0E-12 <5.0E-12 7 HLERF-003 <5.0E-12 <5.0E-12 7 HLERF-004 <5.0E-12 <5.0E-12 7 HLERF-005 <5.0E-12 <5.0E-12 7 HLERF-006 <5.0E-12 <5.0E-12 7 HLERF-007 <5.0E-12 <5.0E-12 7 HLERF-008 <5.0E-12 <5.0E-12 7 HLERF-009 <5.0E-12 <5.0E-12 7 HLERF-010 <5.0E-12 <5.0E-12 7 HLERF-011 1.18E-09 1.18E-09 7 HLERF-012 <5.0E-12 <5.0E-12 7 HLERF-013 <5.0E-12 <5.0E-12 7 HLERF-014 <5.0E-12 <5.0E-12 7 HLERF-015 <5.0E-12 <5.0E-12 7 HLERF-016 <5.0E-12 <5.0E-12 7 HLERF-017 <5.0E-12 <5.0E-12 7 HLERF-018 <5.0E-12 <5.0E-12 7 HLERF-019 <5.0E-12 <5.0E-12 7 HLERF-020 <5.0E-12 <5.0E-12 7 HLERF-028 3.96E-09 6.47E-09 7 HLERF-029 2.55E-09 4.36E-09 7 HLERF-030 6.94E-12 6.96E-12 7 HLERF-038 1.35E-07 1.33E-07 7 HLERF-039 9.15E-08 9.01E-08 7 ILERF-001 2.21E-08 2.19E-08 2

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class INTACT-001 4.26E-07 4.56E-07 1 INTACT-002 2.50E-07 2.51E-07 1 INTACT-003 <5.0E-12 <5.0E-12 1 INTACT-004 <5.0E-12 <5.0E-12 1 INTACT-005 1.50E-06 1.61E-06 1 INTACT-006 1.04E-06 1.04E-06 1 INTACT-007 <5.0E-12 <5.0E-12 1 INTACT-008 <5.0E-12 <5.0E-12 1 INTACT-009 2.36E-08 2.29E-08 1 INTACT-010 <5.0E-12 <5.0E-12 1 INTACT-011 <5.0E-12 <5.0E-12 1 INTACT-012 <5.0E-12 <5.0E-12 1 INTACT-013 8.13E-08 7.86E-08 1 INTACT-014 5.74E-10 5.75E-10 1 INTACT-015 1.13E-10 1.13E-10 1 INTACT-016 <5.0E-12 <5.0E-12 1 INTACT-017 1.12E-07 1.11E-07 1 INTACT-018 6.56E-10 8.85E-10 1 INTACT-019 <5.0E-12 <5.0E-12 1 INTACT-020 <5.0E-12 <5.0E-12 1 INTACT-021 1.62E-06 1.57E-06 1 INTACT-022 4.10E-09 4.08E-09 1 INTACT-023 2.40E-09 2.39E-09 1 INTACT-024 <5.0E-12 <5.0E-12 1 INTACT-028 <5.0E-12 <5.0E-12 1 INTACT-032 <5.0E-12 <5.0E-12 1 INTACT-036 <5.0E-12 <5.0E-12 1 INTACT-040 <5.0E-12 <5.0E-12 1 INTACT-044 <5.0E-12 <5.0E-12 1 INTACT-048 <5.0E-12 <5.0E-12 1

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class INTACT-052 <5.0E-12 <5.0E-12 1 INTACT-056 <5.0E-12 <5.0E-12 1 INTACT-060 <5.0E-12 <5.0E-12 1 INTACT-064 <5.0E-12 <5.0E-12 1 INTACT-068 <5.0E-12 <5.0E-12 1 INTACT-072 <5.0E-12 <5.0E-12 1 LATE-001 <5.0E-12 <5.0E-12 7 LATE-002 1.82E-07 1.78E-07 7 LATE-003 <5.0E-12 <5.0E-12 7 LATE-004 <5.0E-12 2.60E-10 7 LATE-005 <5.0E-12 <5.0E-12 7 LATE-006 <5.0E-12 2.56E-10 7 LATE-007 <5.0E-12 <5.0E-12 7 LATE-008 2.15E-11 1.48E-10 7 LATE-009 <5.0E-12 <5.0E-12 7 LATE-010 6.83E-07 6.55E-07 7 LATE-011 <5.0E-12 <5.0E-12 7 LATE-012 2.99E-07 3.00E-07 7 LATE-013 <5.0E-12 <5.0E-12 7 LATE-014 7.28E-12 2.35E-11 7 LATE-015 <5.0E-12 <5.0E-12 7 LATE-016 <5.0E-12 1.13E-10 7 LATE-017 <5.0E-12 <5.0E-12 7 LATE-018 4.86E-10 1.20E-09 7 LATE-019 <5.0E-12 <5.0E-12 7 LATE-020 <5.0E-12 <5.0E-12 7 LATE-021 <5.0E-12 <5.0E-12 7 LATE-022 1.97E-11 <5.0E-12 7 LATE-023 <5.0E-12 <5.0E-12 7 LATE-024 <5.0E-12 <5.0E-12 7 LATE-025 <5.0E-12 <5.0E-12 7

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class LATE-026 2.12E-09 4.50E-09 7 LATE-027 <5.0E-12 <5.0E-12 7 LATE-028 <5.0E-12 <5.0E-12 7 LATE-029 <5.0E-12 <5.0E-12 7 LATE-030 2.11E-10 4.86E-11 7 LATE-031 <5.0E-12 <5.0E-12 7 LATE-032 <5.0E-12 <5.0E-12 7 LATE-033 <5.0E-12 <5.0E-12 7 LATE-034 1.56E-09 6.26E-10 7 LATE-035 <5.0E-12 <5.0E-12 7 LATE-036 <5.0E-12 <5.0E-12 7 LATE-037 <5.0E-12 <5.0E-12 7 LATE-038 <5.0E-12 <5.0E-12 7 LATE-039 <5.0E-12 <5.0E-12 7 LATE-040 <5.0E-12 <5.0E-12 7 LATE-041 <5.0E-12 <5.0E-12 7 LATE-042 1.33E-06 1.24E-06 7 LATE-043 <5.0E-12 <5.0E-12 7 LATE-044 1.71E-08 1.67E-08 7 LATE-045 <5.0E-12 <5.0E-12 7 LATE-046 3.39E-09 3.51E-09 7 LATE-047 <5.0E-12 <5.0E-12 7 LATE-048 5.80E-09 5.76E-09 7 LATE-055 <5.0E-12 <5.0E-12 7 LATE-056 3.69E-08 6.40E-08 7 LATE-063 <5.0E-12 <5.0E-12 7 LATE-064 2.93E-08 5.06E-08 7 LATE-071 <5.0E-12 <5.0E-12 7 LATE-072 1.44E-08 2.54E-08 7 LATE-079 1.43E-08 2.49E-08 7 LATE-087 <5.0E-12 <5.0E-12 7 LATE-088 9.87E-07 9.72E-07 7

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class LATE-095 <5.0E-12 <5.0E-12 7 LATE-096 8.86E-07 8.73E-07 7 LATE-103 <5.0E-12 <5.0E-12 7 LATE-104 3.13E-06 3.09E-06 7 LATE-111 <5.0E-12 <5.0E-12 7 LATE-112 3.10E-06 3.06E-06 7 LATE-119 <5.0E-12 <5.0E-12 7 LATE-120 7.07E-08 8.48E-08 7 LATE-127 <5.0E-12 <5.0E-12 7 LATE-128 7.13E-08 8.52E-08 7 LATE-135 <5.0E-12 <5.0E-12 7 LATE-136 3.76E-07 3.64E-07 7 LATE-143 <5.0E-12 <5.0E-12 7 LATE-144 3.73E-07 3.62E-07 7 LLERF-001 1.60E-08 1.66E-08 7 LLERF-002 <5.0E-12 <5.0E-12 7 LLERF-003 <5.0E-12 <5.0E-12 7 LLERF-004 <5.0E-12 <5.0E-12 7 LLERF-005 <5.0E-12 <5.0E-12 7 LLERF-006 <5.0E-12 <5.0E-12 7 LLERF-007 5.16E-10 5.18E-10 7 LLERF-008 <5.0E-12 <5.0E-12 7 LLERF-009 <5.0E-12 <5.0E-12 7 LLERF-010 <5.0E-12 <5.0E-12 7 LLERF-011 <5.0E-12 <5.0E-12 7 LLERF-012 <5.0E-12 <5.0E-12 7 LLERF-013 5.56E-10 5.87E-10 7 LLERF-014 <5.0E-12 <5.0E-12 7 LLERF-015 <5.0E-12 <5.0E-12 7 LLERF-016 <5.0E-12 <5.0E-12 7 LLERF-017 <5.0E-12 <5.0E-12 7

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval L2 Accident Unit 1 Unit 2 EPRI Sequence Frequency/yr Frequency/yr Class LLERF-018 <5.0E-12 <5.0E-12 7 LLERF-019 2.64E-08 2.57E-08 7 LLERF-020 4.57E-10 4.58E-10 7 LLERF-021 3.28E-10 3.29E-10 7 LLERF-022 1.76E-11 1.77E-11 7 LLERF-023 1.54E-10 1.54E-10 7 LLERF-024 1.08E-10 1.09E-10 7 LLERF-029 1.21E-10 2.05E-10 7 LLERF-030 8.86E-11 1.34E-10 7 LLERF-035 1.79E-07 1.77E-07 7 LLERF-036 1.49E-07 1.47E-07 7 LLERF-041 2.40E-09 2.81E-09 7 LLERF-042 1.84E-09 2.11E-09 7 LLERF-047 1.78E-08 1.77E-08 7 LLERF-048 1.47E-08 1.46E-08 7 SERF-001 9.94E-08 9.94E-08 6 SERF-002 5.64E-10 5.66E-10 6 SERF-007 <5.0E-12 <5.0E-12 6 SERF-008 <5.0E-12 <5.0E-12 6 SERF-013 7.84E-11 1.08E-10 6 SERF-014 <5.0E-12 <5.0E-12 6 SERF-018 4.51E-08 4.46E-08 6 SERF-019 7.59E-08 7.48E-08 6 SERF-025 1.43E-07 1.41E-07 6 SERF-026 <5.0E-12 <5.0E-12 6 SERF-031 1.70E-09 1.97E-09 6 SERF-032 <5.0E-12 <5.0E-12 6 SERF-037 1.41E-08 1.41E-08 6 SERF-038 <5.0E-12 <5.0E-12 6

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 14 Level 2 Accident Sequence Total Frequency Frequency/yr Accident Sequence Group EPRI Class Unit 1 Unit 2 BLERF 8 6.77E-07 6.68E-07 HLERF 7 2.65E-07 2.68E-07 ILERF 2 2.21E-08 2.19E-08 Intact 1 5.06E-06 5.15E-06 Late 7 1.16E-05 1.15E-05 LLERF 7 4.09E-07 4.06E-07 SERF 6 3.80E-07 3.77E-07 Table 15 presents the EPRI accident class frequencies based on the data from Table 13.Table 15 EPRI Accident Class Frequencies EPRI Accident Class Totals[Table 14]Frequency Classification Unit 1 Unit 2 Sum of EPRI Class 1 5.06E-06 5.15E-06 Sum of EPRI Class 2 2.21E-08 2.19E-08 Sum of EPRI Class 6 3.80E-07 3.77E-07 Sum of EPRI Class 7 1.23E-05 1.21E-05 Sum of EPRI Class 8 6.77E-07 6.68E-07 Sum of LERF Sequences 1.37E-06 1.36E-06 Total Level 2 End-States (Total CDF) Including Class 6 1.84E-05 1.83E-05 Total Level 2 End-States (Baseline CDF) Excluding Class 6 1.80E-05 1.80E-05 7.0 Calculation The section documents the analyses performed for characterizing the effect of containment isolation failures affected by a change in the testing intervals. Section 7 consists of the following sections:Section 7.1 the baseline (3-year CILRT frequency) risk is quantified in terms of frequency per reactor-year for the EPRI accident classes of interest.Section 7.2 the baseline population dose (person-rem) is developed for the applicable accident classes.Section 7.3 the risk impact (in terms of population dose-rate) is evaluated for the EPRI accident classes of interest.Section 7.4 the risk impact in terms of the change in LERF and the change in CCFP is determined.7.1 Step 1 - Baseline Risk Determination Section 7.1 documents the calculations for the quantification of the baseline (3-year CILRT frequency) risk in terms of frequency per reactor year for the EPRI accident classes of interest.[1,§4.2]Section 7.1.1 presents the calculation for the frequency of Class 2 sequences which consist of large containment isolation failures.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Section 7.1.2 presents the calculation for the frequency of Class 7 sequences which consist of early and late severe accident phenomena.Section 7.1.3 presents the calculation for the frequency of Class 8 sequences which consist of bypass events such as a steam generator tube rupture (SGTR) or an unisolable interfacing system LOCA (ISLOCA).Section 7.1.4 presents the calculation for the Type A leakage estimate for the 3a probabilities and frequencies.Section 7.1.5 presents the calculation for the Type A leakage estimate for the 3b probabilities and frequencies.Section 7.1.6 presents the calculation for the Class 1 sequences for the intact containment sequences.7.1.1 Class 2 - Large Containment Isolation Failures This class represents large containment isolation failures. Class 2 contains LERF contributions related to isolation failures without scrubbing credited. The frequency of Class 2 is the sum of those release categories identified in Table 13 as Class 2 taken from the Watts Bar specific Level 2 analysis, and summed in Table 15.Equation 1 Calculation of the Class 2 Frequency Unit 1 FREQClass_2 = Class_2 Accident Sequences

 = 2.21E-08/yr Unit 2 FREQClass_2 = Class_2 Accident Sequences = 2.19E-08/yr 7.1.2 Class 7 - Severe Accident Phenomena Class 7 represents early and late containment failure sequences involving severe accident phenomena related to containment breach and represents contributions to LERF. The frequency of Class 7 is the sum of those release categories identified in Table 13 and summed in Table 15.

Equation 2 Calculation of the Class 7 Frequency Unit 1 FREQClass_7 = Class_7 Accident Sequences

 = 1.23E-05/yr Unit 2 FREQClass_7 = Class_7 Accident Sequences = 1.21E-05/yr Note: Class 7 events contribute to both early and late releases and are distributed as follows:

Unit 1 FREQClass_7_LERF = 6.74E-07/yr FREQClass_7_LATE = 1.16E-05/yr

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Unit 2 FREQClass_7_LERF = 6.74E-07/yr FREQClass_7_LATE = 1.15E-05/yr 7.1.3 Class 8 - Containment Bypass (ISLOCA, SGTR)The frequency of Class 8 is the sum of those release categories identified in Table 13 and summed in Table 15. Class 8 events include ISLOCA events and non-isolable SGTR events (early or late).Equation 3 Calculation of the Class 8 Frequency Unit 1 FREQClass_8 = Class_8 Accident Sequences

 = 6.77E-07/yr Unit 2 FREQClass_8 = Class_8 Accident Sequences = 6.68E-07/yr 7.1.4 Calculation of the 3a Probability and Frequency Containment Type A leakage is associated with EPRI accident Class 3. Consistent with the EPRI methodology[1] Class 3 has been divided into two subclasses, 3a for small liner breaches, and 3b for large liner breaches. The estimate for Class 3 was redistributed back into Class 1 (INTACT).

Therefore each of these classes must be evaluated for applicability to this analysis.The Class 3 containment failures are due to leaks such as liner breaches that could only be detected by performing a Type A CILRT or for egregious cases, by visual inspection. In order to determine the impact of the extended test interval the probability of Type A leakage must be calculated.Calculation of the 3a Probability and Frequency Data presented in the EPRI report[1, §4.1] contains two Type A leakage events out of 217 tests. Using the data, a mean estimate for the probability is determined for Class 3a as shown in Equation 4.Equation 4 Calculation of the Class 3a Probability PClass_3a = #Events ÷ #Tests

 = 2 ÷ 217 = 0.0092 This probability is based on a test interval of 3 tests-in-10 years, as opposed to Watts Bars current 1 test-in-10 years frequency. The probability must be adjusted to reflect this difference which is performed later in this calculation. PClass_3a can be defined as the probability of small pre-existing containment leakage in excess of design allowable (La) but less than 10 La. Probability of 3a is a function of CILRT test interval.

Multiplying the Internal Events (Level 2) baseline CDF by the probability of a Class 3a leak determines the Class 3a frequency contribution in accordance with guidance provided by EPRI.[1]

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Equation 5 Calculation of the Class 3a Failure Frequency Unit 1 FREQClass_3a = PClass_3a x CDFBaseline

 = 0.0092 x 1.80E-05/yr = 1.66E-07/yr Unit 2 FREQClass_3a = PClass_3a x CDFBaseline = 0.0092 x 1.80E-05/yr = 1.66E-07/yr 7.1.5 Calculation of the 3b Probability and Frequency To estimate the failure probability given that no failures have occurred, the EPRI guidance [1, §4.1]

suggests the use of a non-informative prior. This approach updates a uniform distribution (no bias) with the available evidence (data) to provide a better estimation of an event.A beta distribution is typically used for the uniform prior with the parameters = 0.5 and

= 1. This is combined with the existing data (i.e., no Class 3b events in 217 tests) using Equation 6 to calculate the 3b failure probability, and by Equation 7 to calculate the failure frequency.

Equation 6 Calculation of the Class 3b Failure Probability PClass_3b = (n + ) ÷ (N + )

 = (0 + 0.5) ÷ (217 + 1) = 0.5 ÷ 218 = 0.0023 where: n = the number of events of interest (large leakage)

N= the number of tests

 = non-informative prior distribution parameter = non-informative prior distribution parameter PClass_3b can be defined as the probability of large (100 La) pre-existing containment leakage.

Probability of 3b is a function of CILRT test interval.Equation 7 Calculation of the Class 3b Failure Frequency Unit 1 FREQClass_3b = PClass_3b x CDFBaseline

 = 0.0023 x 1.80E-05/yr = 4.14E-08/yr Unit 2 FREQClass_3b = PClass_3b x CDFBaseline = 0.0023 x 1.80E-05/yr = 4.12E-08/yr

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 7.1.6 Class 1 - Intact Containment The Class 1 frequency is determined by the baseline CDF less Class 2, 7, and 8.Equation 8 Calculation of the Class 1 Frequency Unit 1 FREQClass_1 = CDFBaseline - (FREQClass 2 + FREQClass 7 + FREQClass 8)

 = 1.80E-05/yr - (2.21E-08/yr + 1.23E-05/yr + 6.77E-07/yr) = 5.04E-06/yr Unit 2 FREQClass_1 = CDFBaseline - (FREQClass 2 + FREQClass 7 + FREQClass 8) = 1.80E-05/yr - (2.19E-08/yr + 1.21E-05/yr + 6.68E-07/yr) = 5.18E-06/yr Although the frequency of this class is not directly impacted by Type A testing, the frequency for Class 1 is reduced by the estimated frequencies in Class 3a and 3b in order to preserve total baseline CDF. The revised Class 1 frequency is therefore determined by Equation 9:

Equation 9 Calculation of the Adjusted Class 1 Frequency Unit 1 FREQClass 1 ADJ = FREQClass_1_BL - (FREQClass 3a_OLB + FREQClass 3b_OLB)

 = 5.04E-06/yr - (1.66E-07/yr + 4.14E-08/yr) = 4.85E-06/yr Unit 2 FREQClass 1 ADJ = FREQClass 1_BL - (FREQClass 3a_OLB + FREQClass 3b_OLB) = 5.18E-06/yr - (1.66E-07/yr + 4.12E-08/yr) = 4.94E-06/yr 7.2 Step 2 - Develop the Baseline Population Dose In this step, the baseline population dose is calculated. The population dose is a function of the accident class frequency and the population within a 50-mile radius of the Watts Bar plant. The following sub-steps and tables are provided in this section.

Section 7.2.1 presents the Watts Bar 50-mile radius population density Table 16 presents the estimated 2040 population density in the 50-mile radius surrounding Watts Bar Table 17 presents the estimates person-rem for Sequoyah source term groups reported from NUREG/CR-4551.[7] This table also assigns each source term group to an EPRI class Table 18 presents the calculation of the Sequoyah population dose risk at 50-miles by taking the fractional contribution to risk for each collapsed accident progression bin, the population dose risk and the frequency to obtain the estimated dose for each accident progression Section 7.2.2 presents the Watts Bar off-site consequence (person-rem) estimates Table 19 presents the Unit-1 Baseline dose calculation without the breakout of Class 3 which is included in the Class 1 values.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 20 presents the Unit-2 Baseline dose calculation without the breakout of Class 3 which is included in the Class 1 values.Table 21 presents the Unit 1 Baseline Dose with the Class 3a and 3b Contribution Table 22 presents the Unit 2 Baseline Dose with the Class 3a and 3b Contribution The estimated population within a 50-mile radius of the Watts Bar Plant was taken from calculation R-2361441-1823[15 §4.5.2] which was developed for the Unit 2 Severe Accident Mitigation Alternatives (SAMA) Analysis. The basis for the population projections are described fully in that calculation. The results, in part are provided in Table 16.Table 16 50-Mile Radius Population Density Total Population 2040 Data [15, Table 5]1,523,390 Per the EPRI guidance[1 §5.1.2], the population dose is calculated by using the data provided in NUREG/CR-4551 Vol 5 Rev 1 Part 1[7] for the Sequoyah plant and adjusting the results for Watts Bar. Specifically each Watts Bar release category is associated with an applicable collapsed accident progression bin of NUREG/CR-4551. Refer to NUREG/CR-4551 for descriptions of the collapsed accident progression bins (APBs) from NUREG/CR-4551[9, §2.4.3].Table 17 Summary Accident Progression Bin (APB) Descriptions Summary APB Description Number VB, Early CF (During CD) 1 Core damage occurs followed by vessel breach, containment fails early during core damage.VB, Alpha, Early CF (at VB) 2 Core damage occurs followed by vessel breach, steam explosion leads to early containment failure at vessel breach.VB, >200 psi, Early CF (at VB) 3 Core damage occurs leading to a high pressure vessel breach resulting in early containment failure during vessel breach.VB, <200 psi, Early CF (at VB) 4 Core damage occurs, followed by vessel breach at low pressure, containment fails early at vessel breach.VB, Late CF 5 Core damage occurs, followed by vessel breach, containments fails late.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval VB, BMT, Very Late CF 6 Core damage occurs, followed by vessel breach and basemat failure, containments fails very late.Bypass 7 Containment bypass event such as a SGTR.VB, No CF Core damage occurs, followed by vessel breach, pressure is suppressed; therefore, no 8containment failure.No VB, Early CF (during CD) 9 Core damage occurs; however, the vessel remains intact, containment fails early during core damage.No VB 10 Core damage occurs; however, the vessel remains intact and containment does not fail.In order to utilize the Sequoyah information from NUREG/CR-4551 it is necessary to convert the data to the form necessary for the CILRT analysis. This involves classification into one of 3 EPRI classes (2, 7 or 8) and then determining the representative person-rem estimates.NUREG/CR-4551 provides guidance with respect to the composition of the source term grouping. Each source term group is a collection of source terms that result in similar consequences. Therefore, the frequency of the source term group is the sum of the frequencies of all the Accident Progression Bins (APB) which make up the group. Using this information the Sequoyah results are grouped into the EPRI classes and presented in Table 18.Table 18 Calculation of the Watts Bar Population Dose Risk at 50-Miles NUREG/CR-Fractional NUREG/CR-4551 Collapsed 4551 APB Population Dose NUREG/CR-4551 Accident EPRI Collapsed Accident Collapsed Contributions Risk at 50-Miles Population Dose at Progression Class Progression Bin (APB) 1 APB to Risk2 (Person-Rem/yr - 50-Miles6 Bin Frequency/Mean) 4 Yr5 I 7 VB, Early CF (During CD) 0.039 0.468 2.79E-07 1.68E+06 VB, Early CF (At VB), Alpha II 7 0.010 0.114 1.12E-07 1.02E+06 Mode VB >200 psi, Early CF (at III 7 0.159 1.902 1.95E-06 9.74E+05 VB)VB <200 psi, Early CF (at IV 7 0.125 1.494 1.28E-06 1.16E+06 VB)V 7 VB, Late CF 0.073 0.870 2.12E-06 4.10E+05 VI 7 VB, BMT, Very Late CF 0.140 1.674 9.54E-06 1.75E+05 VII 8 Bypass 0.403 4.830 3.12E-06 1.55E+06 VIII 1 VB, No CF, No Bypass 0.002 0.018 1.50E-05 1.20E+03

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval NUREG/CR-Fractional NUREG/CR-4551 Collapsed 4551 APB Population Dose NUREG/CR-4551 Accident EPRI Collapsed Accident Collapsed Contributions Risk at 50-Miles Population Dose at Progression Class Progression Bin (APB) 1 APB to Risk2 (Person-Rem/yr - 50-Miles6 Bin Frequency/Mean) 4 Yr5 No VB, Early CF (During IX 7 0.051 0.612 6.14E-07 9.97E+05 CD)X 1 No VB, No CF 0.002 0.018 2.07E-05 8.69E+02 Total Fractional 1.000 Contribution:

1. The accident progression bins used in the EPRI guidance [1 Table 5-5] were taken from Surry data, the corresponding Watts Bar ABP is used for this analysis, including Vessel Breach

2. This is consistent with the EPRI Guidance. [6 Table 5-5 Note 1] the Fractional APB Contributions to risk is represented by the average of the Mean Fractional Contributions to Risk (MFCR) and the Fractional Contribution to Mean Risk taken from NUREG/CR-4551 Vol 5 Rev 1 Part

1. [7, Table 5.1-3]

3. The total population dose risk (PDR) at 50-miles (PDR50) from internal events is 12.0 person-rem/rx-yr, which is provided as the mean of the sample. [7 Table 5.1-1]

4. The contribution for a given APB is the product of the total PDR50 (12.0 person-rem/rx-yr[7 Table 5.1-1]

 ) and the fractional APB contribution to risk. Multiply column 3 by 12.0. Ex. 0.041

12 = 0.492

5. NUREG/CR-4551 provides the conditional probabilities of the collapsed APBs.[7, Figure 2.5-3] The weighted averages for these conditional probabilities (6th column in Figure 2.5-3, ex. 0.005 for VB, Early CF (During CD)) are multiplied by the Frequency Weighted Average (5.58E-05), to calculate the collapsed APB frequency.

6. The population dose at 50 miles result in this column is determined by dividing the population dose risk shown in the fourth column (0.492) of this table by the collapsed APB (VB, Alpha, Early CF) frequency (2.79E-08) shown in the fifth column of this table, which equals 1.68E+07 Person-Rem at 50 miles.

7.2.1 Watts Bar Specific Off-Site Consequence (Person-Rem Estimates)Watts Bar population dose is calculated using the data provided in NUREG/CR-4551 for Sequoyah and adjusting the results for Watts Bar. Sequoyahs population within a 50 mile radius is 1,224,924 people.[5] A ratio of the population between the two plants is given as:Population of Watts Bar (50 miles) / Population of Sequoyah (50 miles) =1,523,390 / 1,224,924 = 1.24 Population Dose Factor To determine the applicable population dose for Watts Bar, the population dose methodology for the Surry collapsed accident progression bins are referenced. This analysis method from the EPRI guidance [1, Table 5-5] was used to estimate the population density within a 50-mile radius of the Watts Bar plant. The corresponding inputs for Watts Bar are inserted into the equations. The results are provided in Table 19 and Table 20 which present the Watts Bar Unit 1 and Unit 2 release category mapping for the eight EPRI accident classes. Dose (person-rem) per year is the product of the baseline frequency (per year) (from Table 15), the population dose factor from above and the person-rem for an accident class.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Equation 10 Baseline Dose Calculation DoseBL(person-rem/yr) = Baseline Frequency/yr

Dose
Population Dose Factor Table 19 U1 - Baseline Dose Calculation (Without 3a & 3b)

Dose Population EPRI Baseline Person-EPRI Description (Person- Dose Class Frequency/yr Rem/yr Rem) Factor 1 Intact Containment 5.06E-06 2.07E+03 1.24 1.30E-02 Large Containment Isol 2 2.21E-08 9.97E+05 1.24 2.74E-02 Failures Small Isol Failures (Liner 3a (1) (1) 1.24 (1)Breach)Large Isolation failures 3b (1) (1) 1.24 (1)(Liner Breach)Small Isolation Failures -4 (2) (2) 1.24 (2)Failure-to-Seal (Type B)Small Isolation Failures -5 (2) (2) 1.24 (2)Failure-to-Seal (Type C)Containment Isolation 6 Failures (Dependent (2) (2) 1.24 (2)Failure, Personnel Errors)Severe Accident 7 Phenomena Induced 1.23E-05 2.97E+07 1.24 4.55E+02 Failure Containment Bypass 8 6.77E-07 1.55E+06 1.24 1.30E+00 (ISLOCA, SGTR)Total: 1.80E-05 4.559E+02

1. The Class 3a and 3b frequencies, dose and dose-rates are subsumed in the associated Class 1 values.

2. Class 4, 5 and 6 containment isolation failures are not affected by the CILRT interval frequency; therefore they are not included in the analysis.

Table 20 U2 - Baseline Dose Calculation (Without 3a & 3b)Dose Population EPRI Person-EPRI Description Frequency/yr (Person- Dose Class Rem/yr Rem) Factor 1 Intact Containment 5.15E-06 2.07E+03 1.24 1.32E-02 Large Containment 2 2.19E-08 9.97E+05 1.24 2.72E-02 Isolation Failures Small Isolation failures 3a (1) (1) 1.24 (1)(Liner Breach)Large Isolation failures 3b (1) (1) 1.24 (1)(Liner Breach)Small Isolation Failures -4 (2) (2) 1.24 (2)Failure-to-Seal (Type B)Small Isolation Failures -5 (2) (2) 1.24 (2)Failure-to-Seal (Type C)

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Dose Population EPRI Person-EPRI Description Frequency/yr (Person- Dose Class Rem/yr Rem) Factor Containment Isolation Failures (Dependent 6 (2) (2) 1.24 (2)Failure, Personnel Errors)Severe Accident 7 Phenomena Induced 1.21E-05 2.97E+07 1.24 4.47E+02 Failure Containment Bypass 8 6.68E-07 1.55E+06 1.24 1.28E+00 (ISLOCA, SGTR)Total: 1.80E-05 4.485E+01 Table 21 and Table 22 account for the Class 3a and 3b data for Unit 1 and Unit 2, respectively.As such, to conserve CDF, the Class 1 frequency is reduced accordingly.Table 21 U1 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution)Dose Population EPRI Baseline Person-EPRI Description (Person- Dose Class Frequency/yr Rem/yr Rem) Factor 1 Intact Containment(1) 4.85E-06 2.07E+03 1.24 1.25E-02 Large Containment 2 2.21E-08 7.49E+06 1.24 2.06E-01 Isolation Failures Small Isolation failures 3a 1.66E-07 2.07E+04 1.24 4.28E-03 (Liner Breach)Large Isolation failures 3b 4.14E-08 2.07E+05 1.24 1.06E-02 (Liner Breach)Severe Accident 7 Phenomena Induced 1.23E-05 8.63E+05 1.24 1.32E+01 Failure (Early)Containment Bypass 8 6.77E-07 7.14E+05 1.24 6.01E-01 (ISLOCA, SGTR)Total: 1.80E-05 1.404E+01

1. The Class 3a and 3b frequencies, dose and dose-rates were subsumed in the associated Class 1 values. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve CDF.

Table 22 U2 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution)Dose Population EPRI Person-EPRI Description Frequency/yr (Person- Dose Class Rem/yr Rem) Factor 1 Intact Containment(1) 4.94E-06 2.07E+03 1.24 1.27E-02 Large Containment 2 2.19E-08 7.49E+06 1.24 2.04E-01 Isolation Failures

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Dose Population EPRI Person-EPRI Description Frequency/yr (Person- Dose Class Rem/yr Rem) Factor Small Isolation failures 3a 1.66E-07 2.07E+04 1.24 4.26E-03 (Liner Breach)Large Isolation failures 3b 4.12E-08 2.07E+05 1.24 1.06E-02 (Liner Breach)Severe Accident 7 Phenomena Induced 1.21E-05 8.63E+05 1.24 1.30E+01 Failure (Early)Containment Bypass 8 6.68E-07 7.14E+05 1.24 5.93E-01 (ISLOCA, SGTR)Total: 1.80E-05 1.381E+01

7.3 Step 3 - Risk Impact Evaluation In this step, the risk associated with the change in CILRT testing intervals is evaluated in terms of change to the accident class frequencies and population doses for classes 1, 3a and 3b. This is accomplished in a 3 step process.The current surveillance testing requirement of Type A testing and allowed by 10CFR50, Appendix J is at least 1 test-in-10 years based on an acceptable performance history 6 and represents the current licensing basis for Watts Bar. Extending the Type A CILRT interval from 3 tests-in-10 years (original licensing basis) to 1 test-in-10 years (current licensing basis) increased the window of vulnerability for undetected leakage from 18 to 60 months (1/2 the surveillance interval), a factor of 60/18 or a 3.33x increase. Therefore, considering the proposed licensing basis of extending the CILRT Type A test interval from 3 tests-in-10 years to 1 test-in-15 years increases the average time the leaks can be undetected from 18 to 90 months (1/2 the surveillance interval), a factor of 90/18 or a 5x increase.The following sub-steps are contained within Step 3.Section 7.3.1 presents the risk impact for a 1 test-in-10 years test interval Table 23 presents the Unit 1 frequency, dose-rate and dose for a testing interval of 1 test-in-10 years Table 24 presents the Unit 2 frequency, dose-rate and dose for a testing interval of 1 test-in-10 years 6 Defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than 1.0 La.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Section 7.3.2 presents the risk impact for a 1 test-in-15 years test interval Table 25 presents the Unit 1 frequency, dose-rate and dose for a testing interval of 1 test-in-15 years Table 26 presents the Unit 2 frequency, dose-rate and dose for a testing interval of 1 test-in-15 years Section 7.3.3 presents the dose-rate increase and percentile increase Table 27 presents the Unit 1 Class 1 population dose rate increase due to extending the CILRT interval Table 28 presents the Unit 1 Class 3a population dose rate increase due to extending the CILRT interval Table 29 presents the Unit 1 Class 3b population dose rate increase due to extending the CILRT interval Table 30 presents the Unit 1 total population dose rate increase due to extending the CILRT interval Equation 11 calculates the percent increase in total population dose Table 31 presents the Unit 2 Class 1 population dose rate increase due to extending the CILRT interval Table 32 presents the Unit 2 Class 3a population dose rate increase due to extending the CILRT interval Table 33 presents the Unit 2 Class 3b population dose rate increase due to extending the CILRT interval Table 34 presents the Unit 2 total population dose rate increase due to extending the CILRT interval 7.3.1 Risk Impact - 1 Test-in-10 Years Test Interval Based on the approved EPRI methodology and the NEI guidance, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.The risk contribution of 1 test-in-10 years is determined by multiplying the Class 3 accident frequency by a factor of 3.33. Additionally, as the Class 3 sequence are increased, the Class 1 frequency is adjusted downward to maintain the overall core damage frequency constant. The results of this analysis are presented in Table 23 and Table 24 for Units 1 and 2, respectively.Table 23 U1 - Testing Once-in-10 Years Risk Profile Dose Population EPRI Person-Description Freq/yr (Person- Dose Class Rem/yr Rem) Factor 1 Intact Containment(1) 4.35E-06 2.07E+03 1.24 1.12E-02

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 2 Large Cont. Isol. Failures 2.21E-08 7.49E+06 1.24 2.06E-01 Small Isolation Failures (Liner 3a 5.54E-07 2.07E+04 1.24 1.42E-02 Breach) Assumed 10 La Large Isolation Failures (Liner 3b 1.38E-07 2.07E+05 1.24 3.54E-02 Breach) Assumed 100 La Severe Accident Phenomena 7 1.23E-05 8.63E+05 1.24 1.32E+01 Induced Failure (Early & Late)Containment Bypass (ISLOCA, 8 6.77E-07 7.14E+05 1.24 6.01E-01 SGTR)Total: 1.80E-05 1.407E+01

1. The frequency of Class 1 sequences have been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF constant.

Table 24 U2 - Testing Once-in-10 Years Risk Profile EPRI Dose Population Person-Description Freq/yr (Person- Dose Class Rem/yr 1 Intact Containment (1) 4.49E-06 R2.07E+03

 ) F t 1.24 1.16E-02 2 Large Cont. Isolation Failures 2.19E-08 7.49E+06 1.24 2.04E-01 Small Isolation failures (Liner 3a 5.51E-07 2.07E+04 1.24 1.42E-02 Breach)

Large Isolation failures (Liner 3b 1.37E-07 2.07E+05 1.24 3.53E-02 Breach)Severe Accident Phenomena 7 1.21E-05 8.63E+05 1.24 1.30E+01 Induced Failure (Early & Late)Containment Bypass (ISLOCA, 8 6.68E-07 7.14E+05 1.24 5.93E-01 SGTR)Total: 1.80E-05 1.385E+01

1. The frequency of Class 1 scenarios have been reduced by the frequency of Class 3a and Class 3b in order to preserve CDF.

7.3.2 Risk Impact - Once-in-15 Years Test Interval The approach for developing the risk contribution for a 15-yr interval is the same as that used for the 10-yr interval. The increase for a 15-yr CILRT interval is the ratio of the average time for a failure to detect for the increased CILRT test interval (from 18-months to 90-months); therefore the baseline data for Class 3 is multiplied by a factor of 5[1, §4.2.3] and Class 1 is adjusted downward to conserve CDF. The results of this calculation are presented in Table 25 and Table 26 for Units 1 and 2, respectively.Table 25 U1 - Testing Once-in-15 Years Risk Profile Dose Population EPRI Person-Description Freq/yr (Person- Dose Class Rem/yr Rem) Factor 1 Intact Containment(1) 4.00E-06 2.07E+03 1.24 1.03E-02

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Dose Population EPRI Person-Description Freq/yr (Person- Dose Class Rem/yr Rem) Factor Large Containment Isolation 2 2.21E-08 7.49E+06 1.24 2.06E-01 Failures Small Isolation Failures (Liner 3a 8.31E-07 2.07E+04 1.24 2.14E-02 Breach)Large Isolation Failures (Liner 3b 2.07E-07 2.07E+05 1.24 5.32E-02 Breach)Severe Accident Phenomena 7 1.23E-05 8.63E+05 1.24 1.32E+01 Induced Failure (Early & Late)Containment Bypass (ISLOCA, 8 6.77E-07 7.14E+05 1.24 6.01E-01 SGTR)Total: 1.80E-05 1.410E+01

1. The frequency of Class 1 scenarios have been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF constant.

Table 26 U2 - Testing Once-in-15 Years Risk Profile Dose Population EPRI Person-Description Freq/yr (Person- Dose Class Rem/yr Rem) Factor 1 Intact Containment(1) 4.14E-06 2.07E+03 1.24 1.07E-02 Large Containment Isolation 2 2.19E-08 7.49E+06 1.24 2.04E-01 Failures Small Isolation Failures (Liner 3a 8.28E-07 2.07E+04 1.24 2.13E-02 Breach)Large Isolation Failures (Liner 3b 2.06E-07 2.07E+05 1.24 5.30E-02 Breach)Severe Accident Phenomena 7 1.21E-05 8.63E+05 1.24 1.30E+01 Induced Failure (Early & Late)Containment Bypass 8 6.68E-07 7.14E+05 1.24 5.93E-01 (ISLOCA, SGTR)Total: 1.80E-05 1.387E+01

1. The frequency of Class 1 scenarios have been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF constant.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 7.3.3 Dose-Rate Increase and Percentile Increase Given the above estimates, the increases in population dose-rate for each extended interval for EPRI Classes 1, 3a and 3b are estimated as presented in section 7.3.3.1 and 7.3.3.2. Note: The population dose-rate for Class 1 decreases as Class 3b increases to preserve total CDF.Section 7.3.3.1 presents the Unit-1 population dose-rate data and calculations.Section 7.3.3.2 presents the Unit-2 population dose-rate data and calculations.Unit 1 Population Dose-Rate Calculations This section provides the Unit-1 population dose-rate data due to extending the CILRT interval including the OLB comparison with the CLB and PLB intervals, and the difference from CLB and the PLB interval.Table 27 presents the Class 1 population dose-rate increase due to extending the CILRT interval.Table 28 presents the Class 3a population dose-rate increase due to extending the CILRT interval.Table 29 presents the Class 3b population dose-rate increase due to extending the CILRT interval.Table 30 presents the Total (Class 1, 3a and 3b) population dose-rate increase due to extending the CILRT interval.Equation 11 provides the percent increase in population dose-rate calculation.Table 27 U1 - Class 1 PDR Increase Due to Extended Type A CILRT Intervals 3-Yrs 10-Yrs (Baseline - 15-Yrs (Proposed)CILRT Interval (Current)Adjusted) (person-rem)(person-rem)(person-rem)Class 1 PDR 1.25E-02 1.12E-02 1.03E-02 Class 1 PDR (OLBCLB/PLB) -1.30E-03 -2.19E-03 Class 1 PDR (CLBPLB) -8.92E-04 Table 28 U1 - Class 3a PDR Increase Due to Extended Type A CILRT Intervals 3-Yrs 10-Yrs (Baseline - 15-Yrs (Proposed)CILRT Interval (Current)Adjusted) (person-rem)(person-rem)(person-rem)Class 3a PDR 4.28E-03 1.42E-02 2.14E-02 Class 3a PDR (OLBCLB/PLB) 9.97E-03 1.71E-02 Class 3a PDR (CLBPLB) 7.14E-03 Table 29 U1 - Class 3b PDR Increase Due to Extended Type A CILRT Intervals 3-Yrs 10-Yrs (Baseline - 15-Yrs (Proposed)CILRT Interval (Current)Adjusted) (person-rem)(person-rem)(person-rem)

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Class 3b PDR 1.06E-02 3.54E-02 5.32E-02 Class 3b PDR (OLBCLB/PLB) 2.48E-02 4.26E-02 Class 3b PDR (CLBPLB) 1.78E-02 Table 30 U1 -Total PDR Increase Due to Extended Type A CILRT Intervals 3-Yrs 10-Yrs (Baseline - 15-Yrs (Proposed)CILRT Interval (Current)Adjusted) (person-rem)(person-rem)(person-rem)PDRTotal(OLBCLB/PLB)Class 1 + Class 3a + Class 2.74E-02 3.35E-02 5.76E-02 3b PDRTotal(CLBPLB)Class 1 + Class 3a + Class 2.40E-02 3b Given the above values, the percentile increases in total population dose-rate (PDR) for the each test interval compared to the OLB interval are estimated by dividing the increase in total PDR by the baseline total dose for the OLB. The change from the CLB to the PLB uses the total dose for the CLB PDR.Equation 11 Percent Increase in Total Population Dose-Rate (PDR)

 %INCTotal_PDR = (PDR(Change in CILRT Interval) / DOSETotal(CILRT Interval))

100%

Note: PDR(Change in CILRT Interval) taken from Table 30 DOSETotal(CILRT Interval) taken from Tables 21 - 26 Unit-1 Percent Increase in Total PDR (OLBCLB) = (3.35E 2.74E-02 / 14.04)*100%

 = 0.044%

Percentile Increase in Total PDR (CLBPLB) = (5.75E 3.35E-02 / 14.07)*100%

 = 0.171%

Percentile Increase in Total PDR (OLBPLB) = (5.75E 2.74E-02 / 14.10)*100%

 = 0.214%

Unit 2 Population Dose-Rate Calculations The following presents the Unit-2 population dose-rate data due to extending the CILRT interval including the OLB comparison with the CLB and PLB intervals, and the CLB and PLB interval.Table 31 provides the Class 1 population dose-rate increase due to extending the CILRT interval.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 32 provides the Class 3a population dose-rate increase due to extending the CILRT interval.Table 33 provides the Class 3b population dose-rate increase due to extending the CILRT interval.Table 34 provides the Total (Class 1, 3a and 3b) population dose-rate increase due to extending the CILRT interval.Table 31 U2 - Class 1 PDR Increase Due to Extended Type A CILRT Intervals Baseline -10-Yrs (Current) 15-Yrs (Proposed)CILRT Interval Adjusted (person-rem) (person-rem)(person-rem)Class 1 PDR Increase 1.27E-02 1.15E-02 1.05E-02 Class 1 PDR (OLBCLB/PLB) -1.24E-03 -2.13E-03 Class 1 PDR (CLBPLB) -8.89E-04 Table 32 U2 - Class 3a PDR Increase Due to Extended Type A CILRT Intervals Baseline -10-Yrs (Current) 15-Yrs (Proposed)CILRT Interval Adjusted (person-rem) (person-rem)(person-rem)Class 3a PDR Increase 4.26E-03 1.42E-02 2.13E-02 Class 3a PDR (OLBCLB/PLB) 9.93E-03 1.70E-02 Class 3a PDR (CLBPLB) 7.12E-03 Table 33 U2 - Class 3b PDR Increase Due to Extended Type A CILRT Intervals Baseline -10-Yrs (Current) 15-Yrs (Proposed)CILRT Interval Adjusted (person-rem) (person-rem)(person-rem)Class 3b PDR Increase 1.06E-02 3.53E-02 5.30E-02 Class 3b PDR (OLBCLB/PLB) 2.47E-02 4.24E-02 Class 3b PDR (CLBPLB) 1.77E-02 Table 34 U2 Total PDR Increase Due to Extended Type A CILRT Intervals Baseline -10-Yrs (Current) 15-Yrs (Proposed)CILRT Interval Adjusted (person-rem) (person-rem)(person-rem)PDRTotal(OLBCLB/PLB) 2.76E-02 3.34E-02 5.73E-02 Class 1 + Class 3a + Class 3b PDRTotal(CLBPLB) 2.39E-02 Class 1 + Class 3a + Class 3b Given the above values, the percentile increases in total population dose-rate (PDR) for the each test interval compared to the OLB interval are estimated by dividing the increase in total PDR by the baseline total dose for the OLB. The change from the CLB to the PLB uses the total dose for the CLB PDR.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Equation 11 Percent Increase in Total Population Dose-Rate (PDR)

 %INCTotal_PDR = (PDR(Change in CILRT Interval) / DOSETotal(CILRT Interval))

100%

Unit-2 Percent Increase in Total PDR (OLBCLB) = (3.34E 2.76E-02 / 13.81)*100%

 = 0.042%

Percentile Increase in Total PDR (CLBPLB) = (5.73E 3.34E-02 / 13.85)*100%

 = 0.172%

Percentile Increase in Total PDR (OLBPLB) = (5.73E 2.76E-02/ 13.87)*100%

 = 0.214%

7.4 Step 4 - LERF and CCFP Changes In accordance with the methodology presented above, the potential LERF increase due to a CILRT interval extension is estimated as the difference in the Class 3b frequency value of the original licensing basis of 3 tests-in-10 years, the current licensing basis of 1 test-in-10 years, and the proposed licensing basis of 1 test-in-15 years.The risk impact associated with extending the CILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from a containment liner breach could in fact result in a larger release due to the extended window of vulnerability.The additional time between tests could allow for a flaw to continue corroding resulting in a larger containment liner flaw.In accordance with the EPRI guidance, the Class 3a (Small Liner Leak) dose is assumed to be 10 times the allowable intact containment leakage, 10 La (or 2.07E+04 person-rem) and the Class 3b is assumed to be 100 La (or 2.07E+05 person-rem). The method for defining the dose equivalent for allowable leakage (La) is developed in the EPRI report.[1 §4.2.2] The historical observed average leakage is two times La. Therefore, the estimate is conservative.Based on the EPRI guidance, the Class 3 sequences represent those leakage paths that have the potential to result in large release if a pre-existing leak path were present. Class 1 sequences are not considered as potential large release pathways because the containment remains intact.Therefore, the containment leak-rate is expected to be small (i.e., less than 2 La). A larger leak-rate would imply an impaired containment, e.g., Classes 2, 3, 6 and 7. Late releases are excluded regardless of the size of the leak because late releases are by definition, not a LERF event.Therefore, the change in the frequency of Class 3b sequences is used as the increase in LERF for Watts Bar and the change in LERF can be determined by the differences of the 3 test intervals.The EPRI guidance [1 §4.3] states that Class 3b sequences are considered the contributor to LERF associated with the Type A CILRT. Equation 12 for Unit 1 and Unit 2 summarizes the results of the LERF evaluation and the delta LERF for the 3 test intervals.Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. The EPRI guidance cites RG 1.174 and defines small changes in

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval risk as resulting in increases below 1.0E-05/yr and 1.0E-06/yr, for CDF and LERF, respectively.[10§2.4]Since the CILRT frequency does not impact CDF, only LERF is relevant. Furthermore, total LERF from all hazards must be less than 1.0E-05/yr.7.4.1 LERF Determination Equation 12 LERF Determination for Class 3b LERF = FREQClass_3b(CLB or PLB) - FREQClass_3b(OLB)Unit-1 LERF (OLBCLB) = FREQClass_3b (CLB) - FREQClass_3b(OLB)

 = 1.38E-07/yr - 4.14E-08/yr = 9.64E-08/yr LERF (CLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(CLB) = 2.07E-07/yr - 1.38E-07/yr = 6.91E-08/yr LERF (OLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(OLB) = 2.07E-07/yr - 4.14E-08/yr = 1.66E-07/yr Unit-2 LERF (OLBCLB) = FREQClass_3b (CLB) - FREQClass_3b(OLB) = 1.37E-07/yr - 4.12E-08/yr = 9.60E-08/yr LERF (CLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(CLB) = 2.06E-07/yr - 1.37E-07/yr = 6.88E-08/yr LERF (OLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(OLB) = 2.06E-07/yr - 4.12E-08/yr = 1.65E-07/yr 7.4.2 Conditional Containment Failure Probability In accordance with the methodology[1, §5.1.4] presented in Section 5.4.2, the change in the Conditional Containment Failure Probability (CCFP) due to an CILRT interval extension is estimated as the difference in the CCFP values for the original and extended intervals.

Equation 13 Change in CCFP CCFP = [1 - (FREQClass_1(CILRT Interval) + FREQClass_3a(CILRT Interval) / CDFTotal]

100%

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Unit-1 CCFP (OLB) = [1 - (FREQClass_1_U1_ADJ + FREQClass_3a_U1_OLB) / CDFTotal_U1] *100%

 = [1 - (4.85E-06/yr + 1.66E-07/yr) / 1.84E-05/yr] *100% = 72.74%

CCFP (CLB) = [1 - (FREQClass_1_U1_CLB + FREQClass_3a_U1_CLB) / CDFTotal_U1] *100%

 = [1 - (4.35E-06/yr + 5.54E-07/yr) / 1.84E-05/yr] *100% = 73.38%

CCFP (PLB) = [1 - (FREQClass_1_U1_PLB + FREQClass_3a_U1_PLB) / CDFTotal_U1] *100%

 = [1 - (4.00E-06/yr + 8.31E-07/yr) / 1.84E-05/yr] *100% = 73.75%

Unit-2 CCFP (OLB) = [1 - (FREQClass_1_U2_ADJ + FREQClass_3a_U2_OLB) / CDFTotal_U2] *100%

 = [1 - (4.94E-06/yr + 1.66E-07/yr) / 1.84E-05/yr] *100% = 72.16%

CCFP (CLB) = [1 - (FREQClass_1_U2_CLB + FREQClass_3a_U2_CLB) / CDFTotal_U2] *100%

 = [1 - (4.49E-06/yr + 5.51E-07/yr) / 1.84E-05/yr] *100% = 72.52%

CCFP (PLB) = [1 - (FREQClass_1_U2_PLB + FREQClass_3a_U2_PLB) / CDFTotal_U2] *100%

 = [1 - (4.14E-06/yr + 8.28E-07/yr) / 1.84E-05/yr] *100% = 72.89%

Equation 14 %Change CCFP INCCCFP(For given CILRT Interval) = CCFP(CILRT Interval of Interest) - CCFPOLB Unit-1 CCFP Increase (OLBCLB) = CCFPCLB- CCFPOLB

 = 73.37% - 72.73% = 0.638%

CCFP Increase (OLBPLB) = CCFPPLB- CCFPOLB

 = 73.75% - 72.73% = 1.013%

CCFP Increase (CLBPLB) = CCFPPLB- CCFPCLB

 = 73.75% - 73.37% = 0.375%

Unit-2 CCFP Increase (OLBCLB) = CCFPCLB - CCFPOLB

 = 72.58% - 72.23% = 0.354%

CCFP Increase (OLBPLB) = CCFPPLB - CCFPOLB

 = 72.96% - 72.23% = 0.729%

CCFP Increase (CLBPLB) = CCFPPLB- CCFPCLB

 = 72.96% - 72.58% = 0.375%

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 7.4.3 Summary LERF - CCFP Table 35 and Table 36 summarize the LERF and CCFP results for Unit-1 and Unit-2, respectively.Table 35 Unit-1 Summary LERF - CCFP CILRT Interval Risk Metric OLBCLB OLBPLB CLBPLB LERF 9.64E-08 1.66E-07 6.91E-08 CCFP (% Change) 0.638 1.013 0.375 Table 36 Unit-2 Summary LERF - CCFP CILRT Interval Risk Metric OLBCLB OLBPLB CLBPLB LERF 9.60E-08 1.65E-07 6.88E-08 CCFP (% Change) 0.354 0.729 0.375 8.0 Sensitivity Analyses The EPRI guidance for the analysis of extending the CILRT interval suggests using the liner corrosion sensitivity analysis performed by Calvert Cliffs.[12] Additionally the contribution of external events will be addressed in this section. It is important to note that the corrosion analysis is a sensitivity case that represents the first 15-year extension. It is possible for some slow corrosion mechanisms, such as embedment of debris in containment during initial containment construction, where the probability of leakage can continue to increase over longer periods.However, these mechanisms are generally very slow and have a very limited potential for the development of large leakage pathways before detection.[1, §5.1.5.1]8.1 Liner Corrosion The analysis approach uses the Calvert Cliffs Nuclear Plant (CCNP) methodology[12] as modified by the EPRI guidance.[1] This methodology is an acceptable approach to incorporate the liner corrosion issue into the CILRT interval extension evaluation. The results of the analysis indicate that increasing the interval from the original licensing basis to the proposed licensing basis did not significantly increase plant risk of a large early release (Reference Table 40 and Table 41, Unit 1 and 2, respectively).The methodology investigates how an age-related degradation mechanism can be factored into the risk impact associated with longer CILRT testing intervals.The results of the analysis indicate that increasing the interval from 3 years (e.g., 3 tests-in-10 years) to 15 years did not significantly increase plant risk of a large early release.The metric used in the sensitivity analysis is the conditional containment failure probability (CCFP) which is defined as the probability of containment failure given the occurrence of an accident.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The following approach is used to determine the change in likelihood, due to extending the CILRT interval, of detecting corrosion of the steel liner. This likelihood is used to determine the potential change in risk in the form of the sensitivity analysis. Consistent with the Calvert Cliffs analysis, the following are addressed:

Differences between the containment basemat and the containment cylinder and dome,
The historical steel liner flaw likelihood due to corrosion,
The impact of aging,
The corrosion leakage dependency on containment pressure,
The likelihood that visual inspections will be effective at detecting a flaw.

8.1.1 Assumptions Used In the Corrosion Sensitivity Analysis The assumptions used in this sensitivity study are consistent with the Calvert Cliffs methodology and include the following:

1. A half-failure is assumed (e.g., Jeffreys non-informative prior) for basemat concealed liner corrosion due to no operational experience with basemat failures. (Table 37 Step 1)

2. Two corrosion events are used to estimate the liner flaw probability. These events, one at North Anna Unit 2 and the other at Brunswick Unit 2, were initiated from the non-visible (backside) of the containment liner.

3. The success data was limited to 5.5 years to reflect the years since September 1996 when 10CFR50.55a started requiring visual inspection and the Calvert Cliffs analysis. 7 (Table 37 Step 1)

4. The likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists was estimated at 1.0% for the cylinder/dome and 0.1% (1/10 of the cylinder failure probability) for the basemat. These values are conservative as the WBN Level 2 analysis has 1.0% leakage probability at 62 psia; whereas the design pressure is 28.2 psia. [19, §6.2.1.2]

5. The likelihood of leakage escape (due to crack formation) in the basemat region is assumed to be ten times less likely than the containment cylinder and dome region. (Table 37, Step 4)

6. A 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is assumed in the analysis.[6] (Table 37, Step 5)

7. All non-detectable failures are assumed to result in large early releases. This approach is conservative and avoids detailed analysis of containment failure timing and operator recovery actions. That is, the probability of all non-detectable failures from the corrosion sensitivity analysis are added to the EPRI Class 3b (and subtracted from EPRI Class 1).

7 Additional success data was not used to limit the aging impact of the corrosion issue, although inspections were being performed prior to the requirement. Furthermore there was no evidence that other liner corrosion issues were identified.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval

8. The liner flaw likelihood is assumed to double every five years. This is based solely on judgment and was included in the Calvert Cliffs analysis. This is done to address the increased likelihood of corrosion as the liner ages. (Table 37 Steps 2 and 3) 8.1.2 Differences in the Watts Bar Design from Calvert Cliffs Structural Design The WBN containment design[19, §6.2.1.2] complies with NRC General Design Criteria 16 (Containment Design). The primary reactor containment is a freestanding, continuous welded steel membrane structure with a vertical cylinder, hemispherical dome, and a flat circular base. A reinforced concrete shield building, surrounding the steel vessel, allows for collection of any containment leakage into an annular region which is subsequently processed by the Emergency Gas Treatment System (EGTS) before release to the environment. The shield building protects the containment vessel from external events.

The double enclosure concept affords minimal interaction between the containment vessel (leakage barrier) and the reactor building (protected structure); a margin of conservatism in leakage rate from the use of two structures and the EGTS; and a reduction of gaseous and particulate radioactive release due to mixing and holdup prior to filtering and release.

Containment Design - General Information The following information is from Reference [19, §6.2.1.2]

 - Design Pressure, psig 13.5 - Design Temperature 250°F - Design & Maximum Allowable Leakage Rate, 0.25%/day - Free Volume 1,171,012 ft3 [19 Table 6.2.1-13]

Testing The WBN reactor containments meets NRC General Design Criteria (GDC) Criterion 52 and 53 with respect to integrated leak rate testing and inspections.[19, §3.1]
Foundation & Bottom Liner Plate The primary containment foundation consists of a 9-foot thick circular reinforced concrete structural slab. The outer 5 - 13 feet where adjacent to other structures is thickened from 12 - 16 feet. [19, §3.8.5.1.1] The containment bottom liner plate is encased in concrete and is not visible. [19, §3.8.2.2.2]
Visual Accessibility Visual inspections of the containment steel liner to satisfy the applicable requirements of the Technical Specifications and ASME Section XI are governed by procedure.[5)) The purpose of the general visual examination is to detect evidence of abnormal degradation or evidence of structural deterioration that may affect either structural integrity or leak tightness.

The general visual examination is currently performed prior to the ten-year containment CILRT, and during two subsequent outages preceding the next required CILRT.[5))

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The containment liner has areas where visual inspection is impossible to perform. These areas include the containment floor liner which is incased in concrete, the area adjacent to the ice condensers, the fuel transfer tube areas and others.[19, §6.2.1.2]8.1.3 Base Case Risk Assessment Table 37 summarizes the results obtained from the CCNP methodology using plant-specific data for Watts Bar.Table 37 WBN Liner Corrosion Base-Case Risk Assessment Containment Cylinder Containment Step Description and Dome (85%) Basemat (15%)Historical Liner Flaw Likelihood 8 Events: 2 Events: 0 Failure Data: Containment Location (Brunswick & North Anna) Assume a 1/2 Failure 1 2 / (70

5.5) = 5.19E-03 0.5 / (70
5.5) = 1.30E-03 Specific Year Failure Rate Year Failure Rate 1 2.1E-03 1 5.0E-04 Aged Adjusted Liner Flaw Likelihood 9 5-10 (ave) 5.2E-03 5-10 (ave) 1.3E-03 2 15 1.4E-02 15 3.5E-03 15 Year Ave = 6.27E-03 15 Year Ave = 1.57E-03 Range % Increase Range % Increase 1 - 3 yrs 0.71 1 - 3 yrs 0.18 3 Flaw Likelihood at 3, 10 and 15 years. 10 1 - 10 yrs 4.06 1 - 10 yrs 1.02 1 - 15 yrs 11 9.40 1 - 15 yrs 12 2.35 8 Containment location specific (consistent with the Calvert Cliffs analysis)[12]

9 During the 15-year interval, assume the failure rate doubles every five years (14.9% increase per year). The average for the fifth to tenth year set to the historical failure rate. (Consistent with the Calvert Cliffs analysis) 10 Uses age-adjusted liner flaw likelihood (Step 2), assuming failure rate doubles every five years (consistent with Calvert Cliffs) 11 The Calvert Cliffs analysis presents the delta between 3 and 15 years of 8.7% to utilize in the estimation of the delta LERF value. For this analysis however, the values are calculated based on 3, 10 and 15 year intervals, consistent with the desired presentation of the results.[1, §5.2.5.1]12 The Calvert Cliffs analysis presents the delta between 3 and 15 years of 2.2% to utilize in the estimation of the delta-LERF value. For this analysis however, the values are calculated based on 3, 10 and 15 year intervals, consistent with the desired presentation of the results.[1, §5.2.5.1]

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Containment Cylinder Containment Step Description and Dome (85%) Basemat (15%)Likelihood of Breach in Containment Given Liner Flaw 13 41.0% 0.1%Visual Inspection Detection Failure 5 10% 14 100% 15 Likelihood 3-Year (Ave) Interval (OLB) 3-Year (Ave) Interval (OLB) 0.71% x 1.0% x 10% 0.18% x 0.1% x 100%0.00071% 0.00018%Likelihood of Non-Detected Containment Leakage 10-Year Test Interval (CLB) 10-Year Test Interval (CLB) 4.06% x 1.0% x 10% 1.02% x 0.1% x 100%6 (Steps 3 x 4 x 5) 0.00406% 0.00102%15-Year Test Interval (PLB) 15-Year Test Interval (PLB) 9.4% x 1.0% x 10% 2.35% x 0.1% x 100%0.0094% 0.00235%Total Likelihood of Non-Detected OLB: 0.00071% + 0.00018% = 0.00000089% or 8.90E-07 Containment Leakage (Cylinder, Dome CLB: 0.00406% + 0.00102% = 0.00000508% or 5.62E-06 and Basemat) PRB: 0.0094% + 0.00235% = 0.0000094% or 1.18E-05 8.1.4 Likelihood of Non-Detected Containment Leakage and LERF Impact The total likelihood of non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome and the containment basemat. The results from Equation 15 apply to both Watts Bar units.13 The failure probability of the cylinder and dome is assumed to be 1%, and basemat is 0.1% as compared to 1.1% and 0.11% in the Calvert Cliffs analysis.14 5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder but could be detected by CILRT). All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.15 The containment basemat liner cannot be visually inspected.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Equation 15 Total Likelihood of Non-Detected Containment Leakage Total Likelihood of Non-Detected Containment Leakage Non-Det Cont LeakageTotal = Step 6(Cylinder & Dome) + Step 6(Basemat)(OLB) = 0.00071% + 0.00018% = 0.00089% or 8.90E-06 (CLB) = 0.00406% + 0.00102% = 0.00508% or 5.62E-05 (PLB) = 0.00940% + 0.00235% = 0.01175% or 1.18E-04 Equation 16 Liner Corrosion Non-LERF Frequency FREQNon-LERF = CDFBL - FREQClass 2 - FREQClass 3b(OLB) - FREQClass 7_Early - FREQClass 8 Example (U1): = 1.80E-05/yr - 2.21E-08/yr - 4.14E-08/yr - 6.74E-07/yr - 6.77E-07/yr

 = 1.66E-05/yr The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF. The following explains how this data is used in Table 38 and Table 39 Unit-2 Increase in LERF/yr for Unit 1 and 2, respectively.

Per Table 21, the Unit-1 EPRI Class 3b frequency is 4.14E-08/yr.
As shown in Table 15, the Watts Bar CDF associated with accidents that are not independently LERF or could never result in LERF (i.e., SERF, Class 6) is 1.84E-05/yr -

3.80E-07/yr = 1.80E-05/yr

The OLB test interval data determined by Equation 15 (8.90E-06) is multiplied by the non-LERF probability determined by the example in Equation 16 (1.73E-05/yr), which results in an increase in LERF of 1.46E-10/yr (
Table 38) The value represents the increase in the baseline Class 3b frequency due to the corrosion-induced concealed flaw issue.

The term Case denotes the following abbreviations, OLB - Original Licensing Basis, or 1 test every 3 years, CLB - Current Licensing Basis, or 1 test in 10 years, and PLB - Proposed Licensing Basis, or 1 test in 15 years.Equation 17 Liner Corrosion - Increase in LERF Increase In LERF = FREQNon-LERF

PNon-Detected leakage Table 38 Unit-1 Increase in LERF/yr Non- Increase Non-LERF CDF/yr Class Class Class Detected In Case Frequency Baseline 2/yr 3b/yr 8/yr Leakage LERF/yr (Eq. 16)

(Eq. 15) (Eq. 17)

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval OLB 1.80E-05 2.21E-08 4.14E-08 6.77E-07 1.66E-05 8.90E-06 1.48E-10 CLB 1.80E-05 2.21E-08 1.38E-07 6.77E-07 1.66E-05 5.62E-05 9.34E-10 PLB 1.80E-05 2.21E-08 2.07E-07 6.77E-07 1.66E-05 1.18E-04 1.95E-09 Table 39 Unit-2 Increase in LERF/yr Non- Increase Non-LERF CDF/yr Class Class Class Detected In Case Frequency Baseline 2/yr 3b/yr 8/yr Leakage LERF/yr (Eq. 16)(Eq. 15) (Eq. 17)OLB 1.80E-05 2.19E-08 4.12E-08 6.68E-07 1.66E-05 8.90E-06 1.47E-10 CLB 1.80E-05 2.19E-08 1.37E-07 6.68E-07 1.66E-05 5.62E-05 9.31E-10 PLB 1.80E-05 2.19E-08 2.06E-07 6.68E-07 1.66E-05 1.18E-04 1.95E-09 8.1.5 Corrosion Impact on CCFP This section uses the likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists which was estimated at 1.0% for the cylinder/dome and 0.1% (1/10 of the cylinder failure probability) for the basemat. These values are conservative as the WBN Level 2 analysis has 1.0% leakage probability at 62 psia; whereas the design pressure is just 28.2 psia. This methodology is consistent with the Calvert Cliffs methodology.[5]In Equation 18 the increase in the CCFP (From Equation 13) due to assumed corrosion is calculated.Equation 18 Increase in CCFP Due to Increase in Flaw Likelihood INCCCFP(CILRT Interval) = CCFP(CILRT Interval)

1.10% + CCFP(CILRT Interval)

Unit-1 INCCCFP_OLB = CCFPOLB

1.1% + CCFPOLB

 = 0.7273

1.1% + 0.7274

 = 0.7354 INCCCFP_CLB = CCFPCLB

1.1% + CCFPCLB

 = 0.7337

1.1% + 0.7338

 = 0.7419 INCCCFP_PLB = CCFPPLB

1.1% + CCFPPLB

 = 0.7375 *1.1% + 0.7375 = 0.7457 Unit 2 INCCCFP_OLB = CCFPOLB

1.1% + CCFPOLB

 = 0.7223

1.1% + 0.7216

 = 0.7296 INCCCFP_CLB = CCFPCLB

1.1% + CCFPCLB

 = 0.7258

1.1% + 0.7252

 = 0.7332 INCCCFP_PLB = CCFPPLB

1.1% + CCFPPLB

 = 0.7296

1.1% + 0.7289

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval

 = 0.7370 Equation 19 uses the CCFP calculated in Equation 13 and subtracts the Corrosion-Induced CCFP from Equation 18 to determine the increase in the CCFP due to corrosion.

Equation 19 CCFP Increase Due to Corrosion INCCCFP(CILRT Interval) = CCFP(With Corrosion) - CCFP(Without Corrosion)Unit-1 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

 = 0.7354 - 0.7274 = 0.00800 INCCCFP_OLB = CCFPCLB(With) - CCFPCLB(Without) = 0.7419 - 0.7338 = 0.00807 INCCCFP_OLB = CCFPPLB(With) - CCFPPLB(Without) = 0.7457 - 0.7375 = 0.00811 Unit 2 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without) = 0.7296 - 0.7216 = 0.00794 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without) = 0.7332 - 0.7252 = 0.00798 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without) = 0.7370 - 0.7289 = 0.00802 8.1.6 Summary of Base Case and Corrosion Sensitivity Cases The contribution of corrosion-induced LERF likelihood (determined in section 8.1.4) is added to the Class 3b LERF cases and a sensitivity analysis is performed. Table 40 and

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 41 for Unit 1 and Unit 2, respectively, provide a summary of the base case as well as the corrosion sensitivity case. The table is divided into three main columns representing the frequency of the CILRT interval:

Base Case, which represents the Original Licensing Basis (OLB) of 3 tests-in-10 years.
Current Licensing Basis (CLB), 1 test-in-10 years.
Proposed Licensing Basis (PLB), 1 test-in-15 years.

Each of the three columns is sub-divided further into corrosion and non-corrosion cases. For both the corrosion and non-corrosion cases, the frequencies of the EPRI accident classes are provided. In the non-corrosion cases, an additional column titled Person-Rem/Yr is provided.The Person-Rem/Yr column provides the change in person-rem/yr between the corrosion and non-corrosion cases. Negative values in the Person-Rem/Yr column indicate a reduction in the person-rem/yr for the selected accident class. This occurs only in accident Class 1 and is a result of the reduction in the frequency of the accident class 1 and an increase in Class 3b.Rows for the totals, both frequency and dose-rate, are provided in the table. Additional summary rows are also provided, including:

The change in dose-rate, expressed as person-rem/yr and percentage of the total baseline dose is provided in the row below the baseline row.
The Conditional Containment Failure Probability (CCFP) is provided in the next row, followed by the change in the CCFP as a percentage.
Class 3b LERF is provided and the accident Class 3b frequency as well as the change in the Class 3b frequency below.
The row titled LERF Class 3b & Non-Corrosion LERF provides the difference between the non-corrosion and corrosion cases.
The row titled LERF (from base case of 3 per 10 years) provides the change in LERF as a function of CILRT frequency from the OLB. The difference between the non-corrosion and corrosion-cases is provided.
The row titled LERF from 1 per 10 years provides the change in LERF as a function of CILRT frequency from the CLB. The difference between the non-corrosion and corrosion-cases is provided.

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 40 Unit-1 Summary of Base Case and Corrosion Sensitivity Cases Original Licensing Basis Current Licensing Basis (1 per 10 years) Proposed Licensing Basis (1 per 15 years)Base Case (3 per 10 years)EPRI Class Without Corrosion With Corrosion Without Corrosion With Corrosion Without Corrosion With Corrosion Person- Person- Per- Person- Person- Per- Person- Person- Per-Frequency Rem Per Frequency Rem Per Rem Per Frequency Rem Per Frequency Rem Per Rem Per Frequency Rem Per Frequency Rem Per Rem Per Year Year Year Year Year Year Year Year Year 1 4.85E-06 1.25E-02 4.85E-06 1.25E-02 -3.81E-07 4.37E-06 1.12E-02 4.37E-06 1.12E-02 -2.40E-06 4.02E-06 1.04E-02 4.02E-06 1.03E-02 -5.03E-06 2 2.21E-08 2.05E-01 2.21E-08 2.05E-01 N/A 2.21E-08 2.05E-01 2.21E-08 2.05E-01 N/A 2.21E-08 2.05E-01 2.21E-08 2.05E-01 N/A 3a 1.66E-07 4.28E-03 1.66E-07 4.28E-03 N/A 5.54E-07 1.42E-02 5.54E-07 1.42E-02 N/A 8.31E-07 2.14E-02 8.31E-07 2.14E-02 N/A 3b 4.14E-08 1.06E-02 4.15E-08 1.07E-02 3.81E-05 1.38E-07 3.54E-02 1.39E-07 3.57E-02 2.40E-04 2.07E-07 5.32E-02 2.09E-07 5.37E-02 5.03E-04 7 1.23E-05 1.32E+01 1.23E-05 1.32E+01 N/A 1.23E-05 1.32E+01 1.23E-05 1.32E+01 N/A 1.23E-05 1.32E+01 1.23E-05 1.32E+01 N/A 8 6.77E-07 6.01E-01 6.77E-07 6.01E-01 N/A 6.77E-07 6.01E-01 6.77E-07 6.01E-01 N/A 6.77E-07 6.01E-01 6.77E-07 6.01E-01 N/A CDF/Total 1.80E-05 1.40E+01 1.80E-05 1.40E+01 3.77E-05 1.80E-05 1.40E+01 1.80E-05 1.40E+01 2.38E-04 1.80E-05 1.41E+01 1.80E-05 1.41E+01 4.98E-04 Dose Dose Dose Dose Dose Dose (per- 3.35E-02 (per- 3.37E-02 5.76E-02 (per- 5.75E-02 N/A N/A (per-rem/yr)Rate rem/yr) rem/yr) rem/yr)

 %Inc 0.239% %Inc 0.241% %Inc 0.411% %Inc 0.410%

CCFP 72.74% 73.54% 73.26% 74.07% 73.64% 74.45%CCFP N/A N/A 0.523% 0.529% 0.899% 0.908%Class 3b 4.14E-08 4.15E-08 1.38E-07 1.39E-07 2.07E-07 2.09E-07 LERF LERF Class 3b & Non-1.48E-10 N/A 9.34E-10 N/A 1.95E-09 Corrosion LERF LERF (from base case of 3 per 10 years) 9.64E-08 9.72E-08 1.66E-07 1.67E-07 LERF from 1 per 10 years N/A 6.91E-08 7.01E-08

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Watts Bar - PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 41 Unit-2 Summary of Base Case and Corrosion Sensitivity Cases Original Licensing Basis Current Licensing Basis (1 per 10 years) Proposed Licensing Basis (1 per 15 years)Base Case (3 per 10 years)EPRI Without Corrosion With Corrosion Without Corrosion With Corrosion Without Corrosion With Corrosion Class Person- Person- Per- Person- Person- Per- Person- Person- Per-Frequency Rem Per Frequency Rem Per Rem Per Frequency Rem Per Frequency Rem Per Rem Per Frequency Rem Per Frequency Rem Per Rem Per Year Year Year Year Year Year Year Year Year 1 4.94E-06 1.27E-02 4.94E-06 1.27E-02 -3.79E-07 4.46E-06 1.15E-02 4.46E-06 1.12E-02 -2.39E-06 4.11E-06 1.06E-02 4.11E-06 1.03E-02 -5.01E-06 2 2.19E-08 2.04E-01 2.19E-08 2.04E-01 N/A 2.19E-08 2.04E-01 2.19E-08 2.04E-01 N/A 2.19E-08 2.04E-01 2.19E-08 2.04E-01 N/A 3a 1.66E-07 4.26E-03 1.66E-07 4.26E-03 N/A 5.51E-07 1.42E-02 5.51E-07 1.42E-02 N/A 8.28E-07 2.13E-02 8.28E-07 2.13E-02 N/A 3b 4.12E-08 1.06E-02 4.14E-08 1.06E-02 3.79E-05 1.37E-07 3.53E-02 1.38E-07 3.55E-02 2.39E-04 2.06E-07 5.30E-02 2.08E-07 5.35E-02 5.01E-04 7 1.21E-05 1.30E+01 1.21E-05 1.30E+01 N/A 1.21E-05 1.30E+01 1.21E-05 1.30E+01 N/A 1.21E-05 1.30E+01 1.21E-05 1.30E+01 N/A 8 6.68E-07 5.93E-01 6.68E-07 5.93E-01 N/A 6.68E-07 5.93E-01 6.68E-07 5.93E-01 N/A 6.68E-07 5.93E-01 6.68E-07 5.93E-01 N/A CDF/Total 1.80E-05 1.38E+01 1.80E-05 1.38E+01 3.75E-05 1.80E-05 1.39E+01 1.80E-05 1.39E+01 2.37E-04 1.80E-05 1.39E+01 1.80E-05 1.39E+01 4.96E-04 Dose Dose Dose Dose Dose (per- 3.34E-02 (per- 3.34E-02 5.73E-02 (per- 5.76E-02 Dose (per-rem/yr)N/A N/A rem/yr) rem/yr) rem/yr)Rate

 %Inc 0.241% %Inc 0.241% %Inc 0.414% %Inc 0.416%

CCFP 72.16% 72.96% 72.692% 73.49% 73.06% 73.87%CCFP N/A N/A 0.523% 0.529% 0.899% 0.908%Class 3b 4.12E-08 4.14E-08 1.37E-07 1.38E-07 2.06E-07 2.08E-07 LERF LERF Class 3b & Non-1.47E-10 N/A 9.31E-10 N/A 1.95E-09 Corrosion LERF LERF (from base case of 3 per 10 years) 9.60E-08 9.68E-08 1.65E-07 1.67E-07 LERF from 1 per 10 years N/A 6.88E-08 6.98E-08

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval 8.1.7 Liner Corrosion Sensitivity Conclusion As shown in Table 40 and Table 41 for Unit 1 and Unit 2, respectively, the inclusion of corrosion does not result in an increase in LERF sufficient to invalidate the baseline analysis and the overall potential impact is negligible.8.2 Seismic CDF The WBN seismic model has been peer reviewed and the Findings and Observations (F&Os) have been subjected to the Appendix X closure review process, discussed in section 4.5.The seismic PRA CDF value is 2.60E-06/yr and 2.61E-06/yr, for Unit 1 and 2, respectively.[33]9.0 Evaluation of External Events In this step, the potential contribution from external events is estimated as a result of increasing the CILRT interval. Due to lack of detailed Level 2 PRA modeling availability for external events, their potential contribution is limited to a conservative estimate of the change in LERF associated with the CILRT interval extension. External events were evaluated in the Watts Bar 10CFR 50.69 License Amendment Request (LAR) where justification for screening of all other external hazards from further consideration was performed with exception of seismic, internal flooding and internal fires. Both seismic and internal flooding hazards are evaluated with a plant specific PRA model.Internal fires were subjected to the Fire Induced Vulnerability Evaluation (FIVE) in support of the IPEEE submitted to NRC.[5 Fire events were considered to be the most limiting due to their frequency of occurrence and their potential impact on plant operation. Therefore, it is assumed that internal fire events bound the risk contribution from other external events.External events were evaluated in the Watts Bar 10CFR 50.69 License Amendment Request (LAR)[52] where justification for screening of all other external hazards from further consideration was performed with exception of seismic, internal flooding and internal fires. Both seismic and internal flooding hazards are evaluated with plant specific PRA models. Internal fires were subjected to the Fire Induced Vulnerability Evaluation (FIVE) in support of the IPEEE submitted to NRC.[18]9.1 Internal Fires Analysis The findings contained in NUREG-1742[21] indicate that the fire CDF is primarily determined by plant transient type of events such as those from assessed plant transients. The judgment is made based on this observation that it is reasonable to assume that the ratio of intact to impaired containments will be similar for fire as for the internal events such that the total CDF and the breakdown by EPRI Class will be equivalent to that presented for the internal events.The Watts Bar internal fire analysis was originally performed in 1998 for WBN Unit 1 and updated in 2014 for WBN Unit 2,[53] in accordance with the Fire Induced Vulnerability Evaluation (FIVE) approach to meet the requirements for the IPEEE. FIVE is fundamentally a prescriptive fire PRA-based screening approach, which uses progressively more detailed phases of screening. All of the WBN fire areas were screened being lower than 1.0E-06 in Phase II of the analysis.[20]

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval The fire-induced core damage frequencies calculated in Part II of the IPEEE FIVE analysis were not summed to give a total fire-induced core damage frequency because, since all fire areas screened there was not a determination made of the core damage frequency from internal fires.For this analysis, a bounding frequency of 1.0E-06/yr is used.[16, 20] Subsequent to FIVE, the Watts Bar plant has performed modifications to reduce the risk from fire events. These modifications include hot-short probability mitigation and Appendix R modifications (cable routing, cable wrapping, cable tray covers, and others).9.2 Seismic Hazards Analysis Addressed in Section 4.3.4.9.3 Other External Events Analysis The contribution from other external hazards such as high winds, transportation, etc., are relatively insignificant in comparison to the consequences of seismic and internal fire events.Watts Bar screen other external hazards from further consideration as part of the required information for the 10 CFR 50.69 License Amendment Request (LAR).[52] The IPEEE assessment of internal fires concluded that the risk (CDF) is less than 1.0E-06/yr; therefore, 1.0E-06 yr is used in this analysis which bounds internal fire risk. Seismic risk is determined by the SPRA model quantification.9.3.1 External Events Contribution to CDF The following calculation and results are applicable for both Watts Bar units.Equation 20 External Events Contribution to CDF CDFEE-U1 = CDFFire_Ux + CDFSeismic_U1

 = 1.0E-06/yr + 2.60E-06/yr = 3.60E-06/yr CDFEE-U2 = CDFFire + CDFSeismic_U2 = 1.0E-06/yr + 2.61E-06/yr = 3.61E-06/yr Note:

9.3.2 External Events Contribution to LERF In Equation 21 the CDF due to external hazards (internal fire and seismic) risk is multiplied by the probability of a Class 3b event. The product is the external events frequency contribution for Class 3b. This calculation applies to both Unit 1 and Unit 2. The LERF contribution for seismic hazards is taken directly from the SPRA model quantification.Equation 21 External Events Impact on the Class 3b Frequency FREQClass 3b(OLB-EE)U1 = (CDFFire

PClass 3b) + (LERFSeismic_U1
PClass 3b)

 = (1.0E-06/yr

0.002294) + (1.70E-06
002294)

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval

 = 6.19E-09/yr FREQClass 3b(OLB-EE)U2 = (CDFFire

PClass 3b) + (LERFSeismic_U2
PClass 3b)

 = (1.0E-06/yr

0.002294) + 1.73E-06
002294)

 = 6.26E-09/yr 9.3.3 External Events Contribution to LERF for CILRT Interval Extension This section characterizes the change in risk associated with external events. The following tables are provided in this section.

Equation 22 presents the external events impact on the Class 3b frequency for the extended CILRT test intervals Equation 23 presents the external events impact on the change in LERF (3b only) from extended CILRT test intervals Table 42 presents the results for external events contribution to the CILRT interval extensions Equation 24 presents the total external events impact on LERF from extending CILRT test interval Table 43 presents the Unit 1 upper bound on the external events LERF to the CILRT interval extensions Table 44 presents the Unit 2 upper bound on the external events LERF to the CILRT interval extensions In Equation 22 the Class 3b frequency determined by multiplying by the CILRT interval factor, 3.33x for 1 test-in-10 years, and 5.0x for 1 test-in-15 years.Equation 22 EE Impact on the Class 3b Frequencies for Extended CILRT Intervals FREQEE-Class 3b(xx-yr) =FREQClass 3b(OLB-EE)* MULTCLB (or MULTPLB)Unit 1 FREQClass 3b(CLB-EE) = 6.19E-09/yr

3.33

 = 2.06E-08/yr FREQClass 3b(PLB-EE) = 6.19E-09/yr

 = 3.10E-08/yr Unit 2 FREQClass 3b(CLB-EE) = 6.26E-09/yr

3.33

 = 2.09E-08/yr FREQClass 3b(PLB-EE) = 6.26E-09/yr

 = 3.13E-08/yr Given the values calculated by Equation 22, the LERF increases for the extended CILRT intervals due to the external events contribution is determined by Equation 23.

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Equation 23 External Events Contribution to a Change in LERF Unit 1 LERFEE(OLBCLB) = FREQClass 3b(CLB-EE) - FREQClass 3b(OLB-EE)

 = 2.06E-08/yr - 6.19E-09/yr = 1.44E-08/yr LERFEE(OLBPLB) = FREQClass 3b(PLB-EE) - FREQClass 3b(OLB-EE) = 3.10E-08/yr - 6.19E-09/yr = 2.48E-08/yr Unit 2 LERFEE(OLBCLB) = FREQClass 3b(CLB-EE) - FREQClass 3b(OLB-EE) = 2.09E-08/yr - 6.26E-09/yr = 1.46E-08/yr LERFEE(OLBPLB) = FREQClass 3b(PLB-EE) - FREQClass 3b(OLB-EE) = 3.13E-08/yr - 6.26E-09/yr = 2.51E-08/yr Table 42 External Events Contribution to Risk for CILRT Interval Extension FREQEE-LERF/yr LERF Increase/yr Unit OLB-EE CLB-EE PLB-EE OLB-EE CLB-EE OLB-EE PLB-EE 1 6.19E-09 2.06E-08 3.10E-08 1.44E-08 2.48E-08 2 6.26E-09 2.09E-08 3.13E-08 1.46E-08 2.51E-08 Only the Class 3b events are affected by an increase in the CILRT interval. The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in LERF does not exceed the guidance for a small change in risk and does not exceed the 1.0E-6/rx-yr change in LERF. [10, §2.4]

Table 43 U1 - Upper Bound on Class 3b LERF EPRI Accident Classes FREQ/yr LERF Increase/yr Hazard OLB CLB PLB OLB CLB OLB PLB External 16 Events 6.19E-09 2.06E-08 3.10E-08 1.44E-08 2.48E-08 Internal Events 1.41E-06 1.51E-06 1.58E-06 9.64E-08 1.66E-07 Combined 1.42E-06 1.53E-06 1.61E-06 1.11E-07 1.90E-07 16 External Events LERF is determined by adding the contribution from Class 3b for the 3 test intervals, and adding the values to the Class 2 and Class 8 values. Only Class 3b is affected by the CILRT interval, therefore, the values for Class 2 and 8 remain static across the 3 test intervals.

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 44 U2 - Upper Bound on Class 3b LERF EPRI Accident Classes FREQ/yr LERF Increase/yr Hazard OLB CLB PLB OLB CLB OLB PLB External Events 6.26E-09 2.09E-08 3.13E-08 1.46E-08 2.50E-08 Internal Events 1.40E-06 1.50E-06 1.57E-06 9.60E-08 1.65E-07 Combined 1.41E-06 1.52E-06 1.60E-06 1.11E-07 1.90E-07 10.0 Results/Conclusion NEI 94-01, Revision 3-A,[3] describes an NRC-accepted approach for implementing the performance-based requirements of 10CFR50, Appendix J, Option B. It incorporates the regulatory positions stated in R.G. 1.163[9] and includes provisions for permanently extending Type A intervals to 15 years. Based on the results of this analysis, sensitivity studies, and conclusions based on calculations that characterize the change in risk are provided for the extended test intervals in this section. A permanent CILRT Type A extension to 1 test-in-15 years presents an insignificant increase in risk to the general public and plant staff as indicated by the results documented in Table 46.Table 45 presents the figures of merit and acceptance criteria Section 10.1 discussion for the change in LERF Section 10.2 discussion for the change in CCFP Section 10.3 discussion for the change in population dose Table 46 presents the results of the calculations and the applicability to this application Table 45 Acceptance Criteria Acceptance Criteria (Increase Above Figure of Merit Source Baseline)LERF <1.0E-05/rx-yr RG 1.174 §2.4[10]LERF <1.0E-06/rx-yr RG 1.174 §2.4[10]CCFP 17 1.5% NEI 94-01 R3-a §2.2[3]Dose increase 1.0 person-rem/yr or 1% of the total Dose (person-rem) EPRI 1018243 APP H[1]baseline dose, whichever is less restrictive In the discussions that follow the maximum results for Unit 1 and Unit 2 are provided for LERF, CCFP, and Dose. Unit 2 bounds Unit 1 test all 3 figures of merit, therefore only that value is presented in the discussions. However, Table 46 provides the data for both units.10.1 Results Discussion - LERF Regulatory Guide 1.174[10] provides guidance for determining the risk impact of plant specific changes to the licensing basis. Leakage characterized by the Type A test does not affect the Core 17 It should be noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water reactor designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small.[1 §1.2]

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Damage Frequency (CDF). Therefore, there is no change to the plant CDF as a result of implementing this proposed change to the licensing basis. The guidance describes a small change in risk for LERF as less than 1.0E-06/rx-yr, If it can be reasonably shown that the total LERF is less than 1.0E-05/rx-yr. For Watts Bar, the analysis included the estimated contribution from external events in addition to the internal events analysis. Table 43 and Table 44 summarizes the maximum LERF for Watts Bar which is estimated to be 1.66E-07/yr, AND the maximum upper bound total LERF (Including External Events) of 1.90E-07/yr. Both results are within the acceptable bands for a small change in risk according to R.G. 1.174.[10 Figure 4] Table 46 provides the results for Unit 1 and Unit 2 for the internal events LERF, combined external events (EE) and internal events (IE) LERF, and the delta LERF for combined EE and IE results for the change from the original licensing basis (OLB) of 3 tests-in-10 years as compared to the current licensing basis (CLB) of 1 test-in-10 years, and the proposed licensing basis (PLB) of 1 test-in-15 years.10.2 Results Discussion - CCFP In accordance with the methodology in EPRI Report 1018243[1] a maximum conditional containment failure probability (CCFP) increase for Watts Bar from the OLB to the PLB is 0.908%which includes the increased contribution due to aging and corrosion affects. EPRI 1018243 characterizes an increase in the CCFP of 1.5% as very small.[1, §2.2] This is consistent with the NRC Final Safety Evaluation for NEI 94-01 Rev 3-A[3]and EPRI Report 1018243.[1] Therefore, this increase is judged to be small. Table 46 provides the detailed results for Unit 1 and Unit 2 for the CCFP change for the CLB and PLB compared to the OLB.10.3 Results Discussion - Population Dose The proposed licensing change in the Type A CILRT interval to 1 test-in-15 years as measured in terms of an increase on the total integrated plant risk for those accident sequences influenced by Type A testing results in a maximum of 5.76E-02 person-rem/yr which corresponds to increase of 0.41% is calculated for Watts Bar. This value is based on internal events only. EPRI Report 1018243[1] states that a small increase in population dose is defined as 1.0 person-rem/yr or 1% of the total population dose, whichever is less restrictive for the risk impact of the CILRT interval extension to 15 years. Therefore, Watts Bar meets both metrics. This is consistent with the NRC Final Safety Evaluation for NEI 94-01 Rev 3-A[3] and EPRI Report 1018243.[1] Table 46 provides the detailed results for Unit 1 and Unit 2 for the total change in dose and the% change for the CLB and PLB compared to the OLB. All values presented in Table 46 include the increased risk from corrosion.

Calculation No. MDN-000-999-2016-000804 Rev: 000 Plant: WBN Page: 89

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Watts Bar - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A CILRT Interval Table 46 Results Table and Applicability Determination Unit 1 Acceptable for Metric Value Acceptance Criteria Application?LERFIE-Total PLB 1.37E-06/yr

 <1.0E-05/rx-yr Yes LERFTotal(IE & EE) PLB 1.61E-06/yr LERFTotal(OLBCLB) EE&IE 1.11E-07/yr <1.0E-06/rx-yr Yes LERFTotal(OLBPLB) EE&IE 1.90E-07/yr CCFP(OLBCLB), Inc. Corrosion 0.529%

1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.908%DOSE(OLBCLB) 3.37E-02 per-rem/yr DOSE(OLBPLB) 5.75E-02 per-rem/yr <1.0 person-rem/yr or <1% of total Yes

 %DOSE(OLBCLB) 0.24% dose, whichever is less restrictive. %DOSE(OLBPLB) 0.41%

Unit 2 LERFIE-Total PLB 1.36E-06

 <1.0E-05/rx-yr Yes LERFTotal(IE & EE) PLB 1.60E-06 LERFTotal(OLBCLB) EE&IE 1.11E-07 <1.0E-06/rx-yr Yes LERFTotal(OLBPLB) EE&IE 1.90E-07 CCFP(OLBCLB), Inc. Corrosion 0.529%

1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.908%DOSE(OLBCLB) 3.34E-02 per-rem/yr DOSE(OLBPLB) 5.76E-02 per-rem/yr <1.0 person-rem/yr or <1% of total Yes

 %DOSE(OLBCLB) 0.24% dose, whichever is less restrictive. %DOSE(OLBPLB) 0.42%

Enclosure 3 Kalsi Engineering Report 3960C (Proprietary)This document contains information that Kalsi Engineering considers to be proprietary in nature.CNL-20-006

Enclosure 4 Kalsi Engineering Report 3960C (Non-Proprietary)This document does not contain proprietary information.CNL-20-006

Page 1 of24 Evaluation of Higher Test Pressure on Leakage for Watts Bar Document No. 3960C, Rev. 0 Prepared for Tennessee Valley Authority Watts Bar Nuclear Station Spring City, TN Prepared by Reviewed by

 ~: .----: )--f, IG,1-0 2.0 .

Neal Estep, P.E. M. S. ~h.D., P.E. Date QA Approval by Fabiola Rico Date KEI File No.263.92.1 Client Purchase Order No. 6232543 , Rev. 0 Date of Preparation: September 15, 2020

 --- Non-Proprietary Version ---

Kalsi Engineering, Inc. / 745 Park Two Drive / Sugar Land, Texas 77478-2885 / Phone: (281) 240-6500 I Fax: (281) 240-0255

Document 3960C, Rev. 0 Page 2 Revisions Rev. DCR/N Pages Description of Changes No. No. Affected 0 N/A Initial release All Description Pages Main Text 24 Attachment 1 32 Attachment 2 19 Attachment 3 70 Attachment 4 95 Attachment 5 37 Total Pages 277 Non-Proprietary Version

Document 3960C, Rev. 0 Page 3 Table of Contents Page 1 STATEMENT OF CALCULATION PURPOSE 5 1.1 Purpose 5 1.2 Scope 5 1.3 Background 7 2 QUALITY ASSURANCE 8 3 DESIGN METHOD USED 9 3.1 Design Method Description 9 3.1.1 MOV Gate Valves 9 3.1.2 Plug Valves 10 3.1.3 Swing Check Valves 11 3.1.4 Lift/Piston Check Valves 12 3.1.5 Manual and AOV Globe Valves with Metal Diaphragms 12 3.1.6 AOV Globe Valves 13 3.1.7 Solenoid-Operated Globe Valves 13 3.2 Generic Leakage Evaluation 14 4 DESIGN INPUTS 16 4.1 Inputs 16 4.2 Leakage history summary 16 5 ASSUMPTIONS 17 6 RESULTS 18 6.1 Attachment 1: MOV Gate Valves 18 6.2 Attachment 2: MOV/AOV Plug Valves 19 6.3 Attachment 3: Soft-Seated and Hard-Seated Swing Check Valves 19 6.3.1 Soft-Seated Swing Check Valves 19 6.3.2 Metal-Seated Swing Check Valves 20 6.4 Attachment 4: Lift/PISTON Check Valves 20 6.5 Attachment 5: AOV Globe Valves 20 7 CONCLUSIONS AND RECOMMENDATIONS 22 7.1 Conclusions 22 7.2 Recommendations 22 8 REFERENCES 23: MOV Gate Valves: MOV and AOV Plug Valves: Swing Check Valves: Lift/Piston Check Valves AOV Globe Valves Non-Proprietary Version

Document 3960C, Rev. 0 Page 4 List of Tables Description Page Table 1-1: Scope of Valves for Evaluation 6 Table 4-1: Generic Inputs for LLRT Evaluation 16 Table 6-1: Seat Load Reduction for MOV Gate Valves 18 Table 6-2: Seat Contact Force Results for AOV Globe Valves 21 List of Figures Description Page Figure 1-1: Microscopic Flow Path Under Light and Heavy Seating Load [5] 7 Figure 3-1: Illustration of Total Seat Load for MOV Wedge Gate Valves 10 Figure 3-2: Tapered Plug Valve Seat Force vs DP 11 Figure 3-3: Metal Sealed Diaphragm Valve Pressure Balance [14] 13 Figure 3-4: Balanced Disk Solenoid Valve Showing Upstream and Downstream Pressure [11] 14 Non-Proprietary Version

Document 3960C, Rev. 0 Page 5 1STATEMENT OF CALCULATION PURPOSE 1.1 PURPOSE Local leak rate testing (LLRT) is performed to satisfy U.S. Nuclear Regulatory Commission (NRC) regulation 10CFR50 Appendix J [4]1 requirements and is conducted according to ANSI/ANS 56.8, Containment System Leakage Testing Requirements [16].The purpose of this report is to evaluate whether an increase in valve seat leakage is expected for this scope of valves if the Type C LLRT differential pressure were reduced from current levels to a lower level.1.2 SCOPE For this evaluation, the applicable scope are valves where a higher differential pressure results in increased sealing, such as a check valve [16], in the Watts Bar LLRT program. The list of applicable valves and groups is shown in Table 1-1, below.Valves listed as Exclude-A, Exclude-B, and Exclude-C were originally screened in as having an LLRT configuration where test pressure tends to increase seat force. Upon closer examination of these valves, certain design features and operating principles were identified that cause test pressure to provide minimum or compensating seat force. Explanations are provided in Section 3.1.1 Numbers in brackets [ ] are references in Section 8 of this document.Non-Proprietary Version

Document 3960C, Rev. 0 Page 6 Table 1-1: Scope of Valves for Evaluation Attachment Description Valve IDs 1 MOV Wedge Gate Valves 1(2)-FCV-26-240 1(2)-FCV-70-134 Groups 26-2, 62-3, 70-3, 70-4, 1(2)-FCV-26-243 1(2)-FCV-70-87 70-5, and 70-6 1(2)-FCV-62-61 1(2)-FCV-70-90 1(2)-FCV-62-63 2 AOV and MOV Plug Valves 2-FCV-31-305 1(2)-FCV-67-134 Groups 31-2, 67-5, 67-6, and 2-FCV-31-306 1(2)-FCV-67-138 77-2 2-FCV-31-308 1(2)-FCV-67-139 2-FCV-31-309 1(2)-FCV-67-141 2-FCV-31-326 1(2)-FCV-67-142 2-FCV-31-327 1(2)-FCV-67-295 2-FCV-31-329 1(2)-FCV-67-296 2-FCV-31-330 1(2)-FCV-67-297 1(2)-FCV-67-130 1(2)-FCV-67-298 1(2)-FCV-67-131 1(2)-FCV-77-127 1(2)-FCV-67-133 1(2)-FCV-77-128 3 Swing Check Valves 1(2)-CKV-26-1260 1(2)-CKV-67-580C Groups 26-1, 67-2, 70-1, and 1(2)-CKV-26-1296 1(2)-CKV-67-580D 81-1 1(2)-CKV-67-580A 1(2)-CKV-70-679 1(2)-CKV-67-580B 1(2)-CKV-81-502 4 Lift/Piston Check Valves 2-CKV-31-3378 1(2)-CKV-61-680 Groups 31-1, 32-2, 43-1, 61-1, 2-CKV-31-3392 1(2)-CKV-61-692 62-4, 63-1, 67-1, 67-3, and 68- 2-CKV-31-3407 1(2)-CKV-61-745 12-CKV-31-3421 1(2)-CKV-62-639 1-CKV-32-293 1(2)-CKV-63-868 1-CKV-32-303 1(2)-CKV-67-575A 1-CKV-32-313 1(2)-CKV-67-575B 2-CKV-32-323 1(2)-CKV-67-575C 2-CKV-32-333 1(2)-CKV-67-575D 2-CKV-32-343 1(2)-CKV-67-585A 1-CKV-43-834 1(2)-CKV-67-585B 1-CKV-43-841 1(2)-CKV-67-585C 1-CKV-43-883 1(2)-CKV-67-585D 1-CKV-43-884 1(2)-CKV-68-849 1(2)-CKV-61-533 Non-Proprietary Version

Document 3960C, Rev. 0 Page 7 Attachment Description Valve IDs 5 AOV Globe Valves 1-FCV-32-102 1(2)-FCV-63-64 Groups 32-3, 63-2, 63-3, 63-4, 1-FCV-32-110 1(2)-FCV-63-23 63-5, and 68-3 1-FCV-32-80 1(2)-FCV-63-71 2-FCV-32-103 1(2)-FCV-63-84 2-FCV-32-111 1(2)-FCV-68-307 2-FCV-32-81 1(2)-FCV-68-308 Exclude-A Manual and AOV Kerotest 1-BYV-32-288 2-BYV-32-338 See Section Diaphragm Globe Valves 1-BYV-32-298 1(2)-FCV-90-110 3.1.5 for Groups 32-1 and 90-1 1-BYV-32-308 1(2)-FCV-90-111 details 2-BYV-32-318 1(2)-FCV-90-116 2-BYV-32-328 1(2)-FCV-90-117 Exclude-B Manual Dragon Diaphragm 1(2)-ISV-52-500 1(2)-ISV-52-504 See Section Globe Valves 1(2)-ISV-52-501 1(2)-ISV-52-505 3.1.5 for Group 52-1 1(2)-ISV-52-502 1(2)-ISV-52-506 details 1(2)-ISV-52-503 1(2)-ISV-52-507 Exclude-C Target Rock Solenoid Valve 1(2)-FCV-43-202 1-FCV-43-436 See Section Group 43-3 1-FCV-43-208 1-FSV-43-307 3.1.7 for 1(2)-FCV-43-434 1-FSV-43-325 details

1.3 BACKGROUND

Leakage depends on many variables, including seating surface finish (waviness and roughness),seat materials, seat contact width, and seating load. Higher seat loads would tend to decrease the size of microscopic flow passages at the seating interface and diminish the size of the leak passage as shown in the Figure 1-1.Figure 1-1: Microscopic Flow Path Under Light and Heavy Seating Load [5]Non-Proprietary Version

Document 3960C, Rev. 0 Page 8 2QUALITY ASSURANCE This project is performed in accordance with TVA Purchase Order No. 6232543 [2], and the Kalsi Engineering, Inc. (KEI) Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B.The valves within the scope of this evaluation perform a containment isolation function and are safety related.Non-Proprietary Version

Document 3960C, Rev. 0 Page 9 3DESIGN METHOD USED The approach used for gate and globe valves is to evaluate the decrease in seat load resulting from the decrease in LLRT differential pressure to determine if the size of the leak path is increased to such an extent to increase the measured leakage. For tapered plug valves the total sealing force (sum of upstream and downstream seat load) is evaluated within the range of LLRT test pressures.Swing and piston (lift) check valves the approach is to determine the reduction in seating stress for soft-seated designs and the increase in leak path area for metal seated designs. A combination of analysis, industry data, and supporting test data (if available) are used.3.1 DESIGN METHOD DESCRIPTION Valves within the scope of this evaluation were divided into groups based on the analysis approach and are described more fully in Attachments 1 to 5. A summary of the approach for each group is presented in this section.3.1.1 MOV Gate Valves During LLRT, sealing load is applied by mechanical wedging and differential pressure (DP).Mechanical wedging load is a function of the applied force acting on top of the gate (or wedge),the wedge angle, and friction at the disk-to-seat surfaces. A higher applied stem force increases the mechanical wedging load. DP load is proportional to the DP and the area over which the DP acts. A higher DP results in a higher DP applied seat load. For MOVs, the seat load due to mechanical wedging is much greater than that due to the LLRT DP as illustrated in Figure 3-1.The approach for this analysis is to evaluate the reduction in total sealing load due to changing the LLRT test DP from 16.5 psig to 9.0 psig to determine if this reduction is expected to increase the measured leakage.Non-Proprietary Version

Document 3960C, Rev. 0 Page 10 nd D P Me ch a n ical Wedging a ad due to Total Seat Lo Seat Load due to Mechanical Wedging LLRT Range Seat Load u e to LLRT DP Seat Load d 0 0.5 1.0 (Pa)/(LLRT DP)Figure 3-1: Illustration of Total Seat Load for MOV Wedge Gate Valves LLRT DP + Mechanical Wedging Seat load = Total Seat Load 3.1.2 Plug Valves Tapered plug valves have a conical metal plug inserted into a polymer body sleeve with an interference fit to form a seal. As the plug rotates 90° from close to open the plug port aligns with the body sleeve port to form a rectangular flow passage. As the plug rotates 90° back from open to close, the plug port rotates away from the sleeve port leaving the solid portion of the plug wall in contact with the rectangular body sleeve port. Differential pressure applied to the valve increases the seat load on the downstream seat and reduces the upstream seat load, although very slightly over the LLRT test DP range as illustrated in Figure 3-2. However, for the low LLRT test DP the net sealing load between the two seats remains the same such that overall leakage is not expected to increase from the higher to lower LLRT DP conditions.Non-Proprietary Version

Document 3960C, Rev. 0 Page 11 6000 5000 Total Normal Seat Force LLRT DP Range 4000 3000 Seat Load Normal Force 2000 1000 00 50 Valve 100 Differential 150 Pressure 200 250 300 350 400 450 500 Figure 3-2: Tapered Plug Valve Seat Force vs DP Downstream Normal Seat Force = Upstream Normal Seat Force = Total Normal Seat Force 3.1.3 Swing Check Valves There are both soft-seated and metal (or hard)-seated swing check valves within the applicable population for evaluation. Swing check valves rely primarily on differential pressure to form a seal along with some weight (or gravity) force. The approach for soft-seated and metal-seated valves is described below.For the soft-seated swing check valves, the approach is as follows:

1. Determine the peak and average seat contact stresses based on the seat load at the higher and lower LLRT pressures. Verify that the peak seat contact stress is higher than the differential pressure to ensure a positive sealing margin at that pressure.

2. Determine the percent O-ring or soft seal compression. Verify that the sealing load at the higher and lower LLRT pressures provide adequate O-ring compression to ensure sealing.

For the metal-seated valves, the approach is as follows:

1. Determine the percentage increase in the leakage flow area with the decrease in the seat Non-Proprietary Version

Document 3960C, Rev. 0 Page 12 load due to the reduction from the higher to lower LLRT test pressure.

2. A negligible change in the leakage flow area shows that the leakage flow resistance (and corresponding leakage flow coefficient, CL) change is well below 35%. Per Section 3.2, the leakage flow coefficient needs to change greater than 35% to have higher measured leakage at the lower LLRT test pressure.

3.1.4 Lift/Piston Check Valves Lift/Piston check valves have a plug with seat load applied by an internal spring. The spring is sized to provide a certain cracking pressure in the under-seat flow direction. In the opposite flow direction, the spring and differential pressure close the valve and provide a sealing load. Except for Groups 62-4, all lift/piston check valves within the scope of this evaluation have soft seats.Low modulus soft seats require less seat load to form a seal than high modulus metal seats since the material more easily fills small surface asperities that form leak passages. The approach for lift/piston check valves is similar to that used for the swing check valves as described in paragraph 3.1.3.3.1.5 Manual and AOV Globe Valves with Metal Diaphragms All manual globe valves and the System 90 AOVs (Group 90-1) in the scope of this evaluation have dish-shaped metal diaphragms sealing the internal line pressure. Stem packing, if present, acts as a secondary seal. Based on examination of the drawing, pressure applied over-the-seat with the valve in the closed position will act equal and opposite on the plug and diaphragm seal such that the test pressure is effectively balanced and there is no significant change in seat load as a function of LLRT test DP (see Figure 3-3). Based on scaling of the manufacturers drawings [14, 17, 18], the Group 32-1 manual Kerotest and Group 90-1 Kerotest air-operated valves have a metal diaphragm diameter that are 12% to 16% larger than the seat diameter, and the Group 52-1 manual valves metal diaphragm diameter is 28% larger than. Therefore, additional evaluation is not required since higher test pressure will not increase the seat load for these valves.Non-Proprietary Version

Document 3960C, Rev. 0 Page 13 Patm Dish-Shaped Metal Diaphragm Seal Plug Assembly Soft Seat Patm Ptest Figure 3-3: Metal Sealed Diaphragm Valve Pressure Balance [14]Metal Diaphragm DP Force Balances Plug DP Force 3.1.6 AOV Globe Valves Like gate valves, sealing load for AOV globe valves is applied by mechanical wedging and differential pressure (DP). Mechanical wedging load is a function of the applied force acting on top of the plug, the plug angle, and friction at the disk-to-seat surfaces. DP load is a function of the DP and the area over which the DP acts. For AOVs, the seat load due to mechanical wedging is much greater than that due to the LLRT differential pressure.The approach for this analysis is to evaluate the reduction in total sealing load due to decreasing the LLRT test DP to determine if an increase in the measured leakage is expected.3.1.7 Solenoid-Operated Globe Valves Group 43-3 solenoid-operated Target Rock valves are 3/8-inch direct-acting, bi-directional energized-to-open. Per the information provided by TVA, Unit 1 and Unit 2 valves are of welded-bonnet and bolted-bonnet construction, respectively. Since the disc assembly part number in the valve drawings for Unit 1 [12, 13] and Unit 2 [11] is the same (202665-1), the discs between the two Units are identical. The valve drawings show a pressure equalizing port in the disc that connects the under-seat side of the disc to the valve bonnet as shown in Figure 3-4. Using the scaled dimensions from more-legible Unit 2 drawing [11], the scaled seat diameter (0.37) is estimated to be equal to the scaled disc stem diameter (0.37). The scaling accuracy was verified through two independent dimensions on the drawing [11]. This results in a zero unbalanced area for the disc making the seat load independent of the differential pressure (DP) load in either direction (flow under-the-seat or flow over-the-seat). In the valve closed condition, the seat load Non-Proprietary Version

Document 3960C, Rev. 0 Page 14 is applied by the spring preload and the plug weight. Therefore, the seat load will remain the same under the reduced LLRT pressure.Figure 3-4: Balanced Disk Solenoid Valve Showing Upstream and Downstream Pressure[11]3.2 GENERIC LEAKAGE EVALUATION Leakage depends on many variables, including seating surface finish (waviness and roughness),seat materials, seat contact width, and seating load. Fortunately, LLRT valves are proven to meet acceptable leakage criteria under the tested conditions which provides assurance that the seating surfaces are in good condition and that minimum seat loads (lb/in) required to achieve sealing have been obtained.For valves where service pressure will tend to diminish the overall leakage channel opening, as by pressing the disk into or onto the seat with greater force,, ASME OM Code ISTC-3630 [3]allows use of the following formula to estimate the leakage. Ltest, is the measured leakage at a test Non-Proprietary Version

Document 3960C, Rev. 0 Page 15 differential pressure of DPtest, Lmax, is the calculated leakage at DPmax, where DPmax is higher than DPtest.(3-1)

This equation conservatively assumes the microscopic leak flow passages at the seating interface providing the flow resistance remain constant in size with increasing seal load due to a higher-differential pressure condition. It also assumes the leakage is proportional to the square root of the test differential pressure. Higher pressure would tend to decrease the size of microscopic flow passages at the seating interface as shown in Figure 1-1.In this analysis, the seat load decreases with decreasing test differential pressure. The amount of seat load decrease is limited to the portion of seat load produced by differential pressure.For a given leak flow path, the leakage, L, is proportional to the square root of the differential pressure according to ISTC-3630. A leakage coefficient proportionality constant, CL, is given to relate the square root of the pressure differential to the leakage flow rate.

 = (3-2)

Over the small change in test air pressure, any physical properties of the air test fluid associated with CL would be relatively constant (air density, etc.) such that only the change in leak flow passage size is important. Therefore, leakage at Condition 1 and Condition 2 can be determined as follows:1 = 1 1 2 = 2 2 To maintain the same leakage (i.e. L1 = L2) at test DP condition 1 (DP1) and test DP condition 2 (DP2), the following relationship between the leakage coefficient, CL, and DP is obtained:(3-3) 2 1

1 2 Using Equation 3-3, if the LLRT test DP changed from 16.5 psi to 9.0 psi then the leakage coefficient can increase by 35% before the measured leakage at the lower DP2 would increase from the measured leakage at the higher DP1. If the relationship between leakage flow rate and DP were linear or squared (DP2), as suggested by some references [5] for compressible fluid, the predicted allowed change in leakage coefficient would be larger. Therefore, it is conservative to assume a square root relationship.Non-Proprietary Version

Document 3960C, Rev. 0 Page 16 4DESIGN INPUTS 4.1 INPUTS Design inputs specific to each grouping of valves are contained in Attachments 1 to 5. Generic inputs are shown in Table 4-1, below.Table 4-1: Generic Inputs for LLRT Evaluation Parameter Value Reference Maximum LLRT Test Pressure, psig =15 x 1.1 = 16.5 15, 16 15 psig is the maximum containment internal pressure for Watts Bar Calculated peak containment internal pressure related to 9.36 (9.0 used in assessment) 15 the design basis loss-of-cooling accident (LOCA), Pa, psig 4.2 LEAKAGE HISTORY

SUMMARY

Leakage summary for all valves within the scope of this evaluation are provided in the individual Attachments. Some valves show unacceptable leakage history, but it is assumed that corrective actions are performed to restore leakage to acceptable levels (see Assumption 5.2).Non-Proprietary Version

Document 3960C, Rev. 0 Page 17 5ASSUMPTIONS 5.1. Assumptions applicable to each group of valves included in this analysis are identified in Attachments 1 to 5.5.2. For all valves it is assumed that the tested leakage at normal LLRT DP is acceptable. This assumption is reasonable since corrective actions are required if leakage exceeds acceptance criteria and does not require verification.Non-Proprietary Version

Document 3960C, Rev. 0 Page 18 6RESULTS This section provides the analysis results showing that increased measured leakage is not expected from changing LLRT test conditions from 16.5 psig to 9.0 psig.A results summary for each attachment is provided below.6.1 ATTACHMENT 1: MOV GATE VALVES Groups 26-2, 62-3, 70-3, 70-4, 70-5, and 70-6 Seating load reduction due to a decrease in LLRT test pressure from 16.5 psig to 9.0 psig is shown in Table 6-1. As can be seen, all values are bounded by a 6% reduction in seat load and the seat contact force values all exceed 100 lb/in, which is the minimum recommended value for metal seats per References 5 and 6. Also, the corresponding leakage coefficient, CL, is not expected to increase seat leakage at the lower test pressure of 9.0 psig unless the seat contact force decreases by more than 35%.Table 6-1: Seat Load Reduction for MOV Gate Valves Tag Number Seat Contact Force Reduction in Seat Contact Pct (%)at DPtest, lb/in Force at Pa, lb/in Reduction 1-FCV-26-240 246.6 7.6 3.1 2-FCV-26-240 133.1 7.4 5.6 1-FCV-26-243 224.7 7.6 3.4 2-FCV-26-243 127.3 7.4 5.8 1-FCV-62-61 155.2 7.2 4.6 2-FCV-62-61 205.4 7.2 3.5 1-FCV-62-63 137.3 7.1 5.2 2-FCV-62-63 175.1 7.1 4.1 1-FCV-70-87 1107.7 4.4 0.4 2-FCV-70-87 1234.4 4.4 0.4 1-FCV-70-134 143.7 6.1 4.3 2-FCV-70-134 143.4 6.2 4.3 1-FCV-70-90 1136.8 4.4 0.4 2-FCV-70-90 1001.5 4.3 0.4 Non-Proprietary Version

Document 3960C, Rev. 0 Page 19 KEI test results for two wedge gate valves show that 10% is a best available bounding reduction in seat load, above which seat leakage can possibly increase for wedge gate valves. Table 6-1 shows that the total reduction in seat load for all MOVs is less than 6% such that an increase in seat leakage is not expected if the LLRT test pressure were to decrease from 16.5 psi to 9.0 psi.Based on these results, seat leakage for all MOV Gate valves within the scope of this assessment is not expected to increase if tested at a lower differential pressure of 9.0 psig.6.2 ATTACHMENT 2: MOV/AOV PLUG VALVES Groups 31-2, 67-5, 67-6, and 77-2 Total seating torque reduction for MOV plug valves due to a decrease in LLRT test pressure from 16.5 psig to 9.0 psig is 0% due to the seat load sharing between the upstream and downstream seating surfaces.Even though the data to calculate the seat load reduction for the all AOV plug valves is not available, the seat torques for all the plug valves would be independent of the DP load for the range of LLRT test pressures. Therefore, the seat leakage is not expected to increase withing the range of LLRT test pressure.6.3 ATTACHMENT 3: SOFT-SEATED AND HARD-SEATED SWING CHECK VALVES Groups 26-1, 67-2, 70-1, and 81-1 6.3.1 Soft-Seated Swing Check Valves The disk of the valves in Groups 67-2 (Unit 1 and Unit 2) and 70-1 (Unit1) has an O-ring that provides a soft sealing. The percent reduction in the sealing load is 45.5% when the pressure is reduced from 16.5 psig to 9.0 psig. However, the peak and average seat contact stresses are well above the LLRT test pressures. A peak contact stress well above the differential pressure being sealed will ensure a proper sealing. Therefore, the leakage is not expected to increase when the LLRT test pressure is reduced.The valves in Group 70-1 Unit 2 have a resilient seat to provide a sealing at low pressures and a Stellite-6 hard-faced metal seat to provide sealing at higher pressures. The calculation shows that the maximum seal compression occurs at a seat load of approximately 30 lb and develops a corresponding peak seat contact stress of approximately 114 psi. Since the peak seat contact stress of 114 psi is higher than the LLRT test pressures, a good seal is ensured. Also, to reach full seal compression, a seat load of 30 lb is required. However, a seat load of approximately 61 lb is obtained at 9.0 psig pressures which ensures the seal will stay fully compressed to ensure proper sealing. Therefore, the leakage is not expected to increase when the LLRT test pressure is reduced.Non-Proprietary Version

Document 3960C, Rev. 0 Page 20 6.3.2 Metal-Seated Swing Check Valves The valves in Groups 26-1 and 81-1 have a metal-seated disc. The percent reduction in the sealing load is 45.5% when the pressure is reduced from 16.5 psig to 9.0 psig. The reduction in the seat load may increase the leakage coefficient, CL, by increasing the leakage flow area. A model assuming elastic behavior of the surface asperities along the seat contact band is used to predict the change in leakage flow channel area due to a decrease in test pressure. The calculated increase in the leakage flow channel area is less than 0.01% and is therefore not expected to increase the seat leakage when the LLRT test pressure is reduced.6.4 ATTACHMENT 4: LIFT/PISTON CHECK VALVES For soft-seated valves Groups 31-1, 32-1, 61-1, 63-1, 67-1, 67-3, and 68-1, the total seat load reduction due to change in the differential pressure from 16.5 psig to 9.0 psig varies between 34%to 46%. The spring force does not significantly contribute towards the sealing force. Although there is reduction in seat load, the seat contact stress is higher than the LLRT test pressures, which ensures sealing and no expected increase in leakage when the LLRT test pressure is reduced.For the inline check valves in group 43-1, the initial peak contact stress developed between the O-ring and the gland wall will ensure sealing. The differential fluid pressure acting on the O-ring will only increase this contact pressure further. Therefore, the inline check valve will seal with no expected increase in leakage when the LLRT test pressure is reduced.For metal-seated valves in Group 62-4, the calculated increase in leakage channel flow area from 16.5 psig to 9.0 psig is less than 0.02%. Therefore, this small change in leakage channel flow area is not expected to change the leakage coefficient, CL, by more than 35% and leakage is not expected to increase when the LLRT test pressure is reduced.6.5 ATTACHMENT 5: AOV GLOBE VALVES Groups 32-3, 63-2, 63-3, 63-4, 63-5, and 68-3 The sealing load per linear inch for the applicable AOV Globe valves are shown in Table 6-2. The minimum recommended seat contact force value for metal seats is 100 lb/in per Reference 5. The seat contact force values exceed 100 lb/in except for Group 32-3 valves at 16.5 psig and 9.0 psig.Non-Proprietary Version

Document 3960C, Rev. 0 Page 21 Table 6-2: Seat Contact Force Results for AOV Globe Valves Seat Contact Force, lb/in Group ID Component ID Reduction at 9.0 psig At 16.5 psig At 9.0 psig from 16.5 psig 1-FCV-32-80 23.10 18.80 4.30 2-FCV-32-81 1-FCV-32-102 32-3 23.10 18.80 4.30 2-FCV-32-103 1-FCV-32-110 23.10 18.80 4.30 2-FCV-32-111 1-FCV-63-64 63-2 374.16 372.65 1.51 2-FCV-63-64 1-FCV-63-23 63-3 390.20 388.69 1.51 2-FCV-63-23 1-FCV-63-71 63-4 614.57 613.73 0.84 2-FCV-63-71 1-FCV-63-84 63-5 537.24 536.40 0.84 2-FCV-63-84 1-FCV-68-307 146.98 146.75 0.23 2-FCV-68-307 68-3 1-FCV-68-308 146.98 146.75 0.23 2-FCV-68-308 The seat contact force is reduced to 18.61% for Group 32-3 at 9.0 psig. Per Section 3.2, the reduction in seat contact force needs to change the leak path leakage coefficient, CL, by 35% to increase seat leakage at the lower test pressure of 9.0 psig.Based on the simplified but conservative calculation performed in Section 5.3 of Attachment 5, it is expected that the change in the leakage flow area will be less than 0.1% (Table 5-3 in Attachment

5) with the reduction in seat load. Therefore, it is expected that the leakage coefficient, CL, will not increase by 35% which is a threshold for the measured leakage at the lower pressure to increase from the measured leakage at the higher pressure. Therefore, the seat load reduction of 18.61% for Group 32-3 valves is not expected to increase the leakage at pressure Pa. Based on these results, seat leakage for all AOV Globe valves within the scope of this assessment is not expected to increase when the LLRT test pressure is reduced.

Non-Proprietary Version

Document 3960C, Rev. 0 Page 22 7CONCLUSIONS AND RECOMMENDATIONS

7.1 CONCLUSION

S The analysis results show that increased measured leakage is not expected when the LLRT test pressure is reduced to 9.0 psig.7.2 RECOMMENDATIONS Hard-seated check valves in Groups 62-4 (lift/piston) and 26-1, 81-1 (swing) have low seating stresses and a significant decrease in seat load from 16.5 psig to 9 psig test pressures. Although the results based on calculations and engineering judgment show that leakage is not expected to increase for these valves, it is recommended to test one check valve from each group to support the calculation-based conclusions.Non-Proprietary Version

Document 3960C, Rev. 0 Page 23 8REFERENCES

1. Kalsi Engineering, Inc, Kalsi Engineering, Inc. Quality Assurance Manual, Document No. 1500C, Rev. 15, January 2019.

2. TVA Purchase Order No. 6232543, Rev. 0.

3. American Society of Mechanical Engineers, ASME Operation & Maintenance Code OM-2004 Edition through 2006 Addenda.

4. U.S. Nuclear Regulatory Commission, 10CFR50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

5. Instrument Society of America, ISA Handbook of Control Valves, 2nd Edition, 1976.

6. Emerson Automation Solutions, Control Valve Handbook, Fifth Edition, 2017.

7. EPRI 3002008055, Evaluation Guide for Valve Thrust and Torque Requirements., Palo Alto, CA: 2016.

8. KEI Document 2083C, Rev. 0 Grand Gulf Nuclear Station Engineering Report for Disc Bypass Leakage Test and Analysis for 4-inch, 150-pound William Powell Flexible Wedge Gate Valve, September 1999.

9. KEI Document 2116C, Rev. 0 Grand Gulf Nuclear Station Engineering Report for Disc Bypass Leakage Test and Analysis for 4-inch, 900-pound and Analysis for 6-inch, 600-pound William Powell Flexible Wedge Gate Valves, October 2000.

10. TVA Engineering Work Request No. EWR20MEC088032, Generate List of U1 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation, Approval date:

06/09/2020.

11. Target Rock Drawing No. 82KK-003BB, Rev. C, Valve Solenoid Operated Bi-directional Flow Energize to Open 3/8 inch.

12. Target Rock Drawing No. 1015005-3, Rev. B, Sol Oper Valve Assy Bi-Directional Flow Energize to Open Size 1/4 to 1.

13. Target Rock Drawing No. 82KK-003-1, Rev. G, Project Control Drawing Solenoid Oper.

Valve 3/8 Tube Energize to Open.Non-Proprietary Version

Document 3960C, Rev. 0 Page 24

14. Kerotest Drawing TV-D-9909X01S-(2) Rev. A, 2 600 Y-Type Globe Valve w/Soft Seat.

15. Watts Bar UFSAR 6.2, Containment Systems.

16. ANSI/ANS-56.8-1994, American National Standard for Containment System Leakage Testing Requirements.

17. Dragon Valves, Dwg 13824, Rev. C2, Valve, Globe, Packless & Secondary Packing 3/4 Socket Weld, ASME Section III CL 2 ANSI 2500 LB.

18. Flowserve Drawing 10-59621-01, Rev. 0, Y-Type Globe Valve Socket Ends, Stainless Steel Soft Seat w/Air Cylinder, Size: 1.5 Class: 600.

19. TVA Engineering Work Request No. EWR20MEC026076, Generate U2 Containment Isolation Valves List and Design Inputs for Kalsi Engineering Pa impact evaluation, Approval date: 08/19/2020.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 2 Revisions Rev. DCR/N Pages Description of Changes No. No. Affected 0 N/A Initial release All Description Pages Main Text 26 Appendix A 6 Total Pages 32 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 3 Table of Contents Page 1 OBJECTIVE AND SCOPE 5 1.1 Objective 5 1.2 Scope 5 1.3 Historical Leakage 6 2 METHODOLOGY 7 2.1 Variables 7 2.2 Wedge Gate Valve Sealing Load 8 2.3 Mechanical Wedging 8 2.4 Sealing Force due to Differential Pressure 10 2.5 Total Sealing Force and Seal Force Reduction 10 2.6 Change in Seat Leakage 11 3 INPUTS 17 3.1 Calculation Inputs 17 3.1.1 LLTR Test Pressure and Adjusted Maximum Containment Design Pressure 17 4 ASSUMPTIONS 19 5 RESULTS AND CONCLUSION 20 5.1 Seat Load Results 20 5.2 Seat Load Reduction Results 20 5.3 Minimum Seat Thrust Requirements 21 5.4 Maximum LLRT Test Pressure for 10% Seal Load Reduction 22 5.5 Conclusion 22 6 REFERENCES 23 Appendix A - Wedge Gate Valve Leak Testing Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 4 List of Tables Table Description Page Table 1-1: Analysis Scope [3, 17] 5 Table 1-2: Unit 1 and 2 LLRT Leakage History [3, 17] 6 Table 3-1: Common Input Data 17 Table 3-2: Valve-Specific Input Data 18 Table 5-1: Seat Load Results 20 Table 5-2: Seat Load Reduction 21 Table 5-3: Minimum Seating Thrust, FST, for Static MOV Test 21 Table 5-4: Maximum DPtest for 10% Reduction in Seat Load 22 List of Figures Table Description Page Figure 2-1: Sealing Force due to Mechanical Wedging 8 Figure 2-2: Microscopic Flow Path Under Light and Heavy Seating Load [6] 12 Figure 2-3: Change in Seat Load as a Ratio of LLRT test DP (DP1) to Pa (DP2) 13 Figure 2-4: Seat Leakage vs. Reduction in Seating Load Test Data for 4-inch 150 lb Flexible Wedge Gate Valve 15 Figure 2-5: Seat Leakage vs. Reduction in Seating Load Test Data for 4-inch 900 lb Flexible Wedge Gate Valve 16 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 5 1OBJECTIVE AND SCOPE 1.1 OBJECTIVE Kalsi Engineering, Inc. (KEI) has been contracted by Tennessee Valley Authority (TVA) to provide engineering services to evaluate the impact of local leak rate test (LLRT) pressures, DPtest, higher than the calculated peak pressure, Pa, for cases where greater test pressure tends to increase the sealing. This work is being done in accordance with Purchase Order No. 6232543 [2]1.The objective of this report is to determine the impact of the reduced LLRT pressure from the test LLRT DP to Pa on the seat leakage. All work performed under this project was done in accordance with the requirements of the KEI Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B requirements.1.2 SCOPE The scope of this attachment is LLRT motor-operated gate valves. Valve IDs and basic information are shown below in Table 1-1.Table 1-1: Analysis Scope [3, 17]Group Component Id Comp Description Manufacturer 26-2 1,2-FCV-26-240 Reactor Bldg Standpipe Isol Anchor-Darling 26-2 1,2-FCV-26-243 Reactor Coolant Pump Sprinkler Hdr Isol Anchor-Darling 62-3 1,2-FCV-62-61 CVCS Seal Water Return Header Isol W120/Westinghouse Elec 62-3 1,2-FCV-62-63 CVCS Seal Water Return Header Isol W120/Westinghouse Elec 70-3 1-FCV-70-87 Thermal Barrier CCS Return 70-4 2-FCV-70-87 Walworth 70-3 1-FCV-70-134 Thermal Barrier CCS Supply 70-5 2-FCV-70-134 Walworth 70-4 1-FCV-70-90 Walworth: Unit 1 Thermal Barrier CCS Return 70-6 2-FCV-70-90 Flowserve: Unit 2 1Numbers in brackets [ ] are references in Section 6.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 6 1.3 HISTORICAL LEAKAGE Reference 3 provides the LLRT history for the Unit 1 MOVs. Table 1-2, below, summarizes the results.Table 1-2: Unit 1 and 2 LLRT Leakage History [3, 17]Component Id Leakage Results Notes 1-FCV-26-240 Favorable history 2-FCV-26-240 Favorable history None 1-FCV-26-243 Favorable history Unit 2 had high leakage for U2R1, but excellent results for 2-FCV-26-243 Favorable history U2R2.1-FCV-62-61 Unfavorable history 2-FCV-62-61 Favorable History Containment isolation barrier is combined with CKV-62-639 1-FCV-62-63 Favorable history 2-FCV-62-63 Favorable history None 1-FCV-70-87 Favorable history Containment isolation barrier is combined with CKV 2-FCV-70-87 Favorable history 575A. U1C2 outage had high leakage but has been 0 since 1-FCV-70-134 Unfavorable history 2-FCV-70-134 Favorable history Unit 1 unfavorable since U1C10 outage 1-FCV-70-90 Favorable history High leakage for U1C9, but favorable leakage before and 2-FCV-70-90 Favorable history after Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 7 2METHODOLOGY 2.1 VARIABLES Variable Description Units Ao Area based on mean seat diameter =2 in2 4As Stem area at packing = 2 in2 4dm Mean body seat ring diameter [13] in DP Valve differential pressure psi DPtest Differential pressure used for LLRT psi DPtest_10% Test differential that will result in a 10% reduction in sealing load at psi Pa ds Stem diameter at packing [13] in FC Minimum required thrust from MOV set-up sheet or calculation [13] lb FI Inertia overshoot thrust from diagnostic test lb FP Stem rejection force due to internal line pressure lb Fpack Packing friction force from MOV set-up sheet or calculation [13] lb Fs Force applied to the top of the disk by the stem lb FS_DP Sealing load due to differential pressure lb FS_reduction Reduction in seat load due to the difference between DPtest and Pa lb FS_Total Total sealing load lb FST Static sealing thrust = C16(max thrust) - C11(thrust at seat contact) lb FTR Friction force due to torque reaction surface lb Pa Calculated peak containment internal pressure related to the design- psig basis loss-of-coolant accident (LOCA)Pup Valve upstream pressure psig Rr Sealing force lb rt Torque reaction radius = external torque arm radius or mean seat ft radius per Reference 5.SF Stem factor, taken from MOV set-up calculation ft Umeas Thrust measurement uncertainty (percent of reading) %R Reduction in sealing load due to Pa vs. test DP %

 µ Disk-to-seat friction coefficient, close stroke direction [13] -- µt Friction coefficient at torque reaction surface = 0.5 per Reference 5 --

Half wedge angle [13] deg.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 8 2.2 WEDGE GATE VALVE SEALING LOAD During LLRT, sealing load is applied by 1) mechanical wedging and 2) differential pressure (DP).Mechanical wedging load is proportional to the applied force acting on top of the gate (or wedge),the wedge angle, and friction at the disc-to-seat contact surfaces. DP load is proportional to the DP and the area over which the DP acts. For MOVs, the seat load due to mechanical wedging is much greater than that due to the LLRT differential pressure.The approach for this analysis is to:

1. Determine the reduction in total sealing load due to changing the test DP from the current value, Ptest, to the design basis containment pressure, Pa.

2. Determine the value of Ptest required to achieve a 10% reduction in sealing force at Pa. The 10% criterion is based on leak testing and is discussed in Section 2.4.

2.3 MECHANICAL WEDGING Per References 4 and 5, the sealing force, Rr, due to the applied stem force, Fs, on the wedge is determined by Equation 1.The free-body diagram for Equation 1 is shown in Figure 2-1.Figure 2-1: Sealing Force due to Mechanical Wedging Solving for forces in the stem axis direction:

 = 2 (sin + cos ) (1)

Solving for the sealing force, Rr:(2)

2 (sin + cos )Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 9 During motor-operated valve (MOV) diagnostic testing, stem force is typically measured by stem-mounted strain gauges. The force applied to the top of the disk is reduced from the measured stem force by the friction force absorbed by the packing (packing force), the stem blow-out force due to internal valve pressure (stem rejection force), and the stem friction force at the torque reaction surface. Weight terms are small and ignored. Therefore, to determine the force applied to the top of the disk when the valve body is pressurized, the packing force, stem rejection force, and torque reaction force must be subtracted from the total closing stem force, Fc, as follows:Fs = Fc - Fpack - Fp - FTR (3)Since = 2 = , and 4

 = = ( + )

Since the minimum closing force, FC, contains the packing load, Fpack, the subtracted packing load term should also contain the torque reaction factor.Therefore,

 = ( ) (4)

The torque reaction factor (TRF) is determined using the following equation per Reference 5:( )() (5)

 = 1 For LLRT conditions, the downstream pressure will be atmospheric. Therefore, the upstream pressure is equal to the test differential pressure: Pup = DP The minimum closing force, Fc, is taken from the minimum required thrust (Fmin) shown on the MOV set-up sheet [12]. This minimum thrust value must be provided by the actuator for the MOV to perform its design function. 75% of the inertia overshoot load (FI) from diagnostic testing [14]

is added since this load will be present regardless of the minimum thrust setting. Even with variations in FI, this is a conservative approach since MOVs are seldom set to the minimum thrust value.Fc = Fmin + 0.75(FI)The maximum packing force, Fpack, is taken as the maximum packing load shown on the MOV set-up sheet or the MOV calculation. This value cannot be exceeded without evaluation.Three Unit 2 MOVs, 2-FCV-62-061, 2-FCV-62-063, and 2-FCV-70-134 were found to be set below the minimum required thrust values. For these three MOVs, the thrust at switch trip and packing loads are taken from CR-1604752 [9] instead of the set-up sheets or calculations.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 10 The disk-to-seat friction coefficient used in the mechanical wedging equation, µ, is taken from the MOV calculation [13]. These are threshold or bounding coefficients which provide conservative results (i.e. higher friction coefficient results in less sealing force, Rr, for a given Fs).The final equation for the sealing force due to mechanical wedging is:( ) (6)

2 (sin + cos )2.4 SEALING FORCE DUE TO DIFFERENTIAL PRESSURE Differential pressure acts normal to the valve disk surface and produces a sealing force equal to the product of the pressure difference and area over which the DP acts. The area defined by the mean seat ring diameter is used for the area over which the DP acts and the resulting sealing force due to differential pressure is shown in Equation 7._ = (7) 2.5 TOTAL SEALING FORCE AND SEAL FORCE REDUCTION The total sealing force due to mechanical wedging and test differential pressure (DPtest) is:( ) (8)_ = + _ = +2 (sin + cos )Separating static (independent of DP) and dynamic (DP dependent) components:( ) ( ) (9)_ = + ( )2 (sin + cos ) 2 (sin + cos )The reduction in sealing force at the maximum containment accident pressure, Pa, that is lower than an LLRT test DP is determined as follows:( ) (10)_ = [ ] ( )2 (sin + cos )Therefore, the percentage reduction in sealing force due to testing at DPtest which is higher than Pa is determined as follows:_ (11)

 = 100 ( )

Solving Equation 11 to determine the maximum allowable LLRT DPtest such that there is no greater than a 10% reduction in sealing force is determined by the following equation:0.1 1 + ( ) (12)_10% =0.9 ( )Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 11 Where:F1 = ( )Where FC = FST + Fpack B = 2 (sin + cos )For each MOV in the scope of this analysis, the percent reduction, R, in sealing force is determined using Equation 11 and the maximum LLRT test DP that results in a 10% reduction in sealing force is determined using Equation 12.To help Watts Bar establish minimum seating thrust values for static MOV testing, FST, the static seating thrust required to achieve no less than a 10% reduction in sealing force and a minimum seat contact force of 100 lb/in is determined. The maximum static seating thrust from these two calculations becomes the minimum required static seating thrust. Static seating thrust is typically determined from MOV diagnostic testing as the difference between maximum close thrust (C16) and thrust at seat contact (C11) which is essentially F1, shown above.Solving Equation 11 to determine the minimum static seat thrust, FST, required to achieve no greater than a 10% reduction in sealing force is determined by the following equation:( ) (0.9 ) 1 (13)_10% =0.1 Solving Equation 9 to determine the minimum seat load, F1, required to achieve no less than 100 lb/in of seat contact force is determined by the following equation.1 (14)_100/ = [100 ( )]The maximum value from Equation 13 and 14 is presented in Table 5-3 since this represents the limiting static seating thrust required to achieve no greater than a 10% reduction in seat load and at least 100 lb/in of seating contact force.2.6 CHANGE IN SEAT LEAKAGE Leakage depends on many variables, including seating surface finish (waviness and roughness),seat materials, seat contact width, and seating load. Recommended seat loads (pounds force per inch of seat circumference) for metal seated valves range from 100 lb/in. for high pressure drop nearly leak-tight surface [6] to 100 lb/in to 200 lb/in for Class V shutoff with metal seats [7].Fortunately, LLRT valves are proven to meet acceptable leakage criteria under the tested conditions which provides assurance that the seating surfaces are in good condition and that minimum seat loads (lb/in) required to achieve sealing have been obtained.For valves where service pressure will tend to diminish the overall leakage channel opening, as by pressing the disk into or onto the seat with greater force, the ASME OM Code ISTC-3630 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 12[8] allows use of the following formula to estimate the leakage, Lmax, at DPmax given a corresponding leakage, Ltest, and a test differential pressure DPtest (lower than DPmax) as shown in the following formula:(15)

This equation conservatively assumes the microscopic flow passages at the seating interface providing the flow resistance remain constant in size with increasing seal load due to a higher-differential pressure condition. It also assumes the leakage is proportional to the square root of the test differential pressure. Higher pressure would tend to decrease the size of microscopic flow passages at the seating interface as shown in Figure 2-2.Figure 2-2: Microscopic Flow Path Under Light and Heavy Seating Load [6]In this analysis, the seat load decreases with decreasing test differential pressure. The amount of seat load decrease is limited to the portion of seat load produced by differential pressure. For an MOV, the total seat load will change as the new test differential pressure, DP2, is reduced from the LLRT test pressure, DP1 as shown in Figure 2-3. However, the seat load due to mechanical wedging is typically much greater than that due to differential pressure such that total seat load change with decreasing differential pressure is small.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 13 dD P nical Wedging an S ea t L oa d d ue to Mecha Total Seat Load due to Mechanical Wedging LLRT Range Seat Load u e to LLRT DP Seat Load d 0 0.5 1.0 (Pa)/(LLRT DP)Figure 2-3: Change in Seat Load as a Ratio of LLRT test DP (DP1) to Pa (DP2)For a given leak flow path, the leakage, L, is proportional to the square root of the differential pressure according to ISTC-3630. A leakage coefficient proportionality constant, CL, is given to relate the square root of the pressure differential to the leakage flow rate.

 = (16)

Over the small change in test air pressure, any physical properties associated with CL would be relatively constant (density, etc.) such that only the change in flow passage size is of importance.Therefore, leakage at Condition 1 and Condition 2 can be determined as follows:1 = 1 1 2 = 2 2 To maintain the same leakage at test DP condition 1 (DP1) and test DP condition 2 (DP2) by setting L1 = L2, and the following relationship between the leakage coefficient, CL, and DP is obtained from Equation 13:2 1 (17)

1 2 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 14 Using Equation 14, if the test DP changed from 16.5 psi to 9 psi then the leakage coefficient needs to increase by 35% before the measured leakage at the lower DP2 would increase from the measured leakage at the higher DP1. If the relationship between leakage flow rate and DP were linear or squared, as suggested by some references [6] for compressible flow, the predicted allowed change in leakage coefficient would be larger. Therefore, it is conservative to assume a square root relationship.To determine the expected change in seat leakage coefficient, CL, with seat load, seat leakage testing performed by KEI [10, 11] for a 4-inch, 150 lb and a 4-inch, 900-lb Powell Flexible Wedge gate valve are examined. This testing showed the expected increase in seat leakage as the seat load was decreased. In this testing, the valves were closed to different stem thrust values to produce mechanical wedging and then differential pressure was applied in the opposite direction to reduce the seat load. The differential pressure was increased until high leakage was obtained. Although the test DP is acting in the opposite direction to that simulated during the LLRT, this testing does provide an indication of the leakage as a function of seat load. A plot of seat leakage versus decrease in seat load is shown in Figure 2-4 for the 150 lb class valve and Figure 2-5 for the 900 lb class valve. The initial seat load was at a high enough level where the leakage was zero or near zero as in the case for the LLRT. As can be seen from this figure, decreases in seat load of 10% or less for the 150 lb class valve and 15% or less for the 900 lb class valve produced very little increase, if any, in seat leakage.Based on this testing, a conservative approach is to assume that the leak path constant, CL, is inversely proportional to the seat load such that an X% decrease in seat load results in an X%increase in CL up to a bounding 10% reduction in seat load based on test data for the 150 lb class valve. It is then possible to show that the decrease in seat load due to reduced DP conditions would not increase the measured seat leakage.As described in the previous example, if the test DP change from 16.5 psi to 9 psi resulted in a 10% reduction in seat load, the leak path constant, CL, could increase by 35% before the enlarged leak path is offset by the decrease in differential pressure and the measured seat leakage would increase.The approach can be summarized as follows:

1. Determine the seat load produced by mechanical wedging and LLRT test DP at 16.5 psi and the portion produced by the test DP.

2. Determine the reduction in seat load for a change in test pressure from 16.5 psi to 9.0 psi.

3. If the reduction in seat load is less than 10% then, based on previous testing, the leakage is not expected to increase. 10% is also considered to be conservative since, per Equation 14, Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 15 the leak path leakage coefficient, CL, could change by Sqrt(16.5/9) = 1.35 (35% increase) before the leakage is expected to increase.

4. Determine the theoretical LLRT test DP that would result in a 10% reduction in seat load at Pa = 9.0 psi.

5. Determine the minimum mechanical seating thrust required to achieve no greater than a 10% reduction in seating load or a minimum seat contact load of 100 lb/in, whichever thrust is greater.

2500 2000 Seat Lekage, ml 1500 1000 500 00% 10% 20% 30% 40% 50% 60%Decrease in Seat Load Figure 2-4: Seat Leakage vs. Reduction in Seating Load Test Data for 4-inch 150 lb Flexible Wedge Gate Valve (Note: Different colors represent different closing wedge thrust loads)Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 16 1000 900 800 700 Seat Leakage, ml 600 500 400 300 200 100 00% 20% 40% 60% 80% 100%Decrease in Seat Load Figure 2-5: Seat Leakage vs. Reduction in Seating Load Test Data for 4-inch 900 lb Flexible Wedge Gate Valve (Note: Different colors represent different closing wedge thrust loads)Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 17 3INPUTS 3.1 CALCULATION INPUTS The input data for the analyses are documented in Table 3-1 and Table 3-2 and were obtained from calculations, drawings, specifications, and other information provided by TVA. Some of these inputs are provided in Appendix B. Inputs that require additional clarification are documented below, if applicable. Justified assumptions were made where data were not available. It is important to note that the results of this analysis may be significantly affected by changing key inputs. It will be necessary to perform an impact analysis if key data are changed in the future.Table 3-1: Common Input Data Item Variable Value Reference Area based on mean seat diameter Ao Calculated Section 2.1 Stem area at packing As Calculated Section 2.1 LLRT test differential pressure, psi DPtest 16.5 See 3.1.1 Adj. calculated peak containment internal Pa 9.0 See 3.1.1 pressure during LOCA, psig Torque reaction radius, ft rt Calculated Section 2.1 Friction coefficient at torque reaction surface µt 0.5 [5]3.1.1 LLTR Test Pressure and Adjusted Maximum Containment Design Pressure The maximum permissible LLRT test pressure is 1.1 x 15= 16.5 psi [15, 16]. The calculated peak containment internal pressure during a loss-of-coolant accident (LOCA) for Watts Bar is 9.36 psig[15]. For purposes of this analysis, a lower and more conservative value of 9 psig is used.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 18 Table 3-2: Valve-Specific Input Data Valve ID Input Parameter [12, 13, 14]FC Fpack FI ds Dm µ SF lb lb lb in in deg ft 1-FCV-26-240 2716 1000 3306 1.0 4.25 5 0.57 0.0148 2-FCV-26-240 3004 1250 438 1.0 4.15 5 0.57 0.0148 1-FCV-26-243 2640 1000 2873 1.0 4.25 5 0.57 0.0148 2-FCV-26-243 2640 1000 452 1.0 4.15 5 0.57 0.0148 1-FCV-62-61 3624 1250 645 1.250 4.064 7 0.6582 0.0116 2-FCV-62-61Note 1 4290 1200 1062 1.250 4.064 7 0.6582 0.0116 1-FCV-62-63 3095 1250 489 1.250 4.064 7 0.57 0.0116 2-FCV-62-63 Note 1 3847 1555 810 1.250 4.064 7 0.57 0.0116 1-FCV-70-87 13158 1000 1338 1.0 2.625 5 0.6 0.0126 2-FCV-70-87 13158 1000 3362 1.0 2.625 5 0.6 0.0126 1-FCV-70-134 2000 875 1298 0.875 3.422 5 0.6393 .0097 2-FCV-70-134 Note 1 2108 762 1098 0.875 3.45 5 0.641 0.0170 1-FCV-70-90 13158 1000 1803 1.0 2.625 5 0.6 0.0126 2-FCV-70-90 12710 1750 1243 1.125 2.625 5 0.6 0.0126 Note 1: Three Unit 2 MOVs, 2-FCV-62-061, 2-FCV-62-063, and 2-FCV-70-134 were found to be set below the minimum required thrust values. For these three MOVs, the thrust at switch trip and packing loads are taken from CR-1604752 [9] instead of the set-up sheets or calculations.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 19 4ASSUMPTIONS Data that have not been formally verified are treated as assumptions. Where possible, the basis of the data has been noted. The following general assumptions were used in this analysis.

1. Seat leakage is proportional to the square root of differential pressure for purposes of determining the allowable increase in the seat leakage coefficient. Use of this relationship in lieu of a linear or squared relationship for this purpose is shown to be conservative in Section 2.6 and does not require verification.

2. The weight of stem and disk are neglected in determining the stem forces acting to push the disk into the seat. Since the weight term will generally provide additional closing force, this is a conservative assumption and does not require verification.

3. 75% of the inertia overshoot load is credited to provide additional stem force after actuator switch-off and is taken from the latest static diagnostic test data. Future variation in inertia overshoot will occur but is reasonably bounded by using the 75% reduction, minimum required thrust from the MOV set-up sheet, and the maximum packing force.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 20 5RESULTS AND CONCLUSION 5.1 SEAT LOAD RESULTS Seating load results for the applicable MOV Gate valves are shown below. As can be seen, the seat load contribution from the test differential is a small contribution to the total seat load for most MOVs.Table 5-1: Seat Load Results Tag Number Seat Contact Force Seat Contact Force due to DP Seat Contact Force at DPtest, lb/in DP, lb/in as a Percent of Total 1-FCV-26-240 246.6 17.5 7.1 2-FCV-26-240 133.1 17.1 12.9 1-FCV-26-243 224.7 17.5 7.8 2-FCV-26-243 127.3 17.1 13.5 1-FCV-62-61 155.2 16.8 10.8 2-FCV-62-61 205.4 16.8 8.2 1-FCV-62-63 137.3 16.8 12.2 2-FCV-62-63 175.1 16.8 9.6 1-FCV-70-87 1107.7 10.8 1.0 2-FCV-70-87 1234.4 10.8 0.9 1-FCV-70-134 143.7 14.1 9.8 2-FCV-70-134 143.4 14.2 9.9 1-FCV-70-90 1136.8 10.8 1.0 2-FCV-70-90 1001.5 10.8 1.1 5.2 SEAT LOAD REDUCTION RESULTS Seating load reduction due to a decrease in LLRT test pressure from the normal test DPtest to Pa (16.5 psig to 9 psig) is shown in Table 5-2. As can be seen, all values are bounded by a 6%reduction in seat load and the seat contact force values all exceed 100 lb/in, which is the minimum recommended value for metal seats per References 6 and 7. Also, the corresponding leakage coefficient, CL, is not expected to increase seat leakage at the lower test pressure of 9 psig unless the seat contact force decreased by more than 35%.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 21 Table 5-2: Seat Load Reduction Tag Number Seat Contact Force Reduction in Seat Contact Pct at DPtest, lb/in Force at Pa, lb/in Reduction 1-FCV-26-240 246.6 7.6 3.1 2-FCV-26-240 133.1 7.4 5.6 1-FCV-26-243 224.7 7.6 3.4 2-FCV-26-243 127.3 7.4 5.8 1-FCV-62-61 155.2 7.2 4.6 2-FCV-62-61 205.4 7.2 3.5 1-FCV-62-63 137.3 7.1 5.2 2-FCV-62-63 175.1 7.1 4.1 1-FCV-70-87 1107.7 4.4 0.4 2-FCV-70-87 1234.4 4.4 0.4 1-FCV-70-134 143.7 6.1 4.3 2-FCV-70-134 143.4 6.2 4.3 1-FCV-70-90 1136.8 4.4 0.4 2-FCV-70-90 1001.5 4.3 0.4 5.3 MINIMUM SEAT THRUST REQUIREMENTS Table 5-3 presents the minimum seat thrust required to maintain no more than a 10% reduction in seat load and 100 lb/in seat contact force using Equations 13 and 14. For MOVs, the seat load is typically determined by the difference between static diagnostic trace markers C16 (maximum thrust) and C11 (thrust at seat contact). The fourth column is the maximum of columns 2 and 3.Table 5-3: Minimum Seating Thrust, FST, for Static MOV Test Tag Number Seat Thrust to achieve Seat Thrust to achieve Minimum 10% Reduction in Seat 100 lb/in of Seat Contact Required Contact Force, lb. Force, lb. Thrust, lb.1-FCV-26-240 1087 1519 1519 2-FCV-26-240 1035 1492 1492 1-FCV-26-243 1087 1519 1519 2-FCV-26-243 1035 1492 1492 1-FCV-62-61 1144 1727 1727 2-FCV-62-61 1144 1727 1727 1-FCV-62-63 1006 1534 1534 2-FCV-62-63 1006 1534 1534 1-FCV-70-87 411 1083 1083 2-FCV-70-87 411 1083 1083 1-FCV-70-134 770 1394 1394 2-FCV-70-134 806 1445 1445 1-FCV-70-90 411 1083 1083 2-FCV-70-90 398 1086 1086 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 22 5.4 MAXIMUM LLRT TEST PRESSURE FOR 10% SEAL LOAD REDUCTION The maximum LLRT test pressure (DPtest) to ensure no greater than a 10% seal load reduction at Pa is shown in Table 5-4. As can be seen, the bounding (lowest) test pressure is 22.1 psi, which exceeds the current maximum LLRT pressure of 16.5 psi.Table 5-4: Maximum DPtest for 10% Reduction in Seat Load Tag Number DPtest 1-FCV-26-240 34.8 2-FCV-26-240 22.7 1-FCV-26-243 32.4 2-FCV-26-243 22.1 1-FCV-62-61 26.0 2-FCV-62-61 31.8 1-FCV-62-63 24.0 2-FCV-62-63 28.4 1-FCV-70-87 217.0 2-FCV-70-87 241.0 1-FCV-70-134 27.5 2-FCV-70-134 27.1 1-FCV-70-90 222.5 2-FCV-70-90 202.3

5.5 CONCLUSION

To be at risk for increased leakage at a Pa = 9.0 psi, the seat load would need to decrease by 35%.KEI test results show that 10% is a best available bounding reduction in seat load, above which seat leakage can increase for wedge gate valves. Table 5.1 shows that many of the MOVs have seat load contributions due to DP that are less than 10%. Table 5.2 shows that the total reduction in seat load for all MOVs is less than 6% such that an increase in seat leakage is not expected if the LLRT test pressure were to decrease from 16.5 psi to 9 psi. Table 5-3 provides the minimum seat thrust from MOV static diagnostic testing to achieve no greater than a 10% reduction in seating load or a minimum seat contact load of 100 lb/in, whichever thrust is greater. Table 5.4 shows that the limiting maximum LLRT pressure for all MOVs such that the seat load would not drop by more than 10% is 22.1 psi.Based on these results, seat leakage for all MOV Gate valves within the scope of this assessment is not expected to increase if tested at a lower differential pressure of 9.0 psig.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 23 6REFERENCES

1. KEI Document No. 1500C Rev. 15; Kalsi Engineering, Inc. Quality Assurance Manual.

2. TVA Purchase Order 6232543, Rev. Num: 0.

3. TVA Engineering Work Request, EWR20MECH088032, Generate List of U1 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 06/09/20.

4. Evaluation Guide for Valve Thrust and Torque Requirements. EPRI, Palo Alto, CA: 2016.

3002008055.

6. ISA Handbook of Control Valves, 2nd Edition, J.W. Hutchison, Instrument Society of America, 1979.

7. Control Valve Handbook, Fifth Edition, Emerson Automation Solutions, 2017.

8. ASME OM-2004 Edition through 2006 Addenda, Operation and Maintenance of Nuclear Power Plants.

9. TVA Watts Bar, White Paper for Appendix J Leak Tightness for Program MOVs Using Close Torque Switch Bypass Motor Control Set at no Less than 98%, CR 1604752, WF Cetta, 4/29/2020, Rev. 1, 07/06/2020.

10. KEI Document 2083C, Rev. 0 Grand Gulf Nuclear Station Engineering Report for Disc Bypass Leakage Test and Analysis for 4-inch, 150-pound William Powell Flexible Wedge Gate Valve, September 1999.

11. KEI Document 2116C, Rev. 0 Grand Gulf Nuclear Station Engineering Report for Disc Bypass Leakage Test and Analysis for 4-inch, 900-pound and Analysis for 6-inch, 600-pound William Powell Flexible Wedge Gate Valves, October 2000.

12. MOV Set-Point Sheets

a. 1-FCV-26-240-A, 1-47A8910-26-07, R5 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 24

b. 1-FCV-26-243-A, 1-47A8910-26-10, R5

c. 2-FCV-26-240-A, 2-47A8910-26-07, R4

d. 2-FCV-26-243-A, 2-47A8910-26-10, R2

e. 1-FCV-62-61-B, 1-47A8910-62-01, R3

f. 1-FCV-62-63-B, 1-47A8910-62-2, R4

g. 2-FCV-62-61-B, 2-47A8910-62-01, R2

h. 2-FCV-62-63-B, 2-47A8910-62-02, R2

i. 1-FCV-70-87-B, 1-47A8910-70-03, R3

j. 1-FCV-70-134-B, 1-47A8910-70-09, R2

k. 2-FCV-70-87-B, 2-47A8910-70-03, R2

l. 2-FCV-70-134-B, 2-47A8910-70-09, R3

m. 1-FCV-70-90-A, 1-47A8910-70-05, R3

n. 2-FCV-70-90-A, 2-47A8910-70-05, R3

13. MOV Set-up Calculations

a. TVA Calculation T71191007812, Rev. 009, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-26-240.

b. TVA Calculation T93150924008, Rev. 004, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-26-0240.

c. TVA Calculation T71191007813, Rev. 010, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-26-243.

d. TVA Calculation T93150807008, Rev. 003, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-26-0243.

e. TVA Calculation T71030218808, Rev. 5, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-62-61.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 25

f. TVA Calculation T93150112024, Rev. 005, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-62-061.

g. TVA Calculation T71110212844, Rev. 006, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-62-63.

h. TVA Calculation T93140404009, Rev. 003, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-62-063.

i. TVA Calculation T71191007811, Rev. 010, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-70-87.

j. TVA Calculation T71191007816, Rev. 002, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-70-087.

k. TVA Calculation T71030613801, Rev. 9, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-70-90.

l. TVA Calculation T93140426049, Rev. 003, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-70-090.

m. TVA Calculation T71191007810, Rev. 007, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 1-FCV-70-134.

n. TVA Calculation T71191007818, Rev. 008, Documentation of Design Basis Review, Required Thrust/Torque Calculations and Valve and Actuator Capability Assessment for Valve 2-FCV-70-134.

14. Diagnostic Test Data

a. WID: 120295525, VOTES Infinity Valve Diagnostic System Report, WBN FCV-026-0240-A, 5/17/2020.

b. WID: 120295526, VOTES Infinity Valve Diagnostic System Report, WBN FCV-026-0243-A, 5/23/2020.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Page 26

c. WID: 111457730, MOVATS Testing of Motor Operated Valves, 2-FCV-026-0240, 7/27/2015.

d. WID: 111457734, MOVATS Testing of Motor Operated Valves, 2-FCV-026-0243, 7/20/2015.

e. WID: 119083315, MOVATS Testing of Motor Operated Valves, 1-MVOP-062-0061-B, As-Left, 9/29/2018.

f. WID: 118305103, MOVATS Testing of Motor Operated Valves, 1-FCV-062-0063, As-Left, 3/4/17.

g. WID: 118771402, MOVATS Testing of Motor Operated Valves, 2-FCV-62-61-B, As-Left, 11/04/2017.

h. WID: 112216193, MOVATS Testing of Motor Operated Valves, 2-FCV-062-0063-A, 2/17/14.

i. WID: 112800521, MOVATS Testing of Motor Operated Valves, 1-FCV-70-87, As-Left, 9/15/12.

j. WID: 115752256, MOVATS Testing of Motor Operated Valves, 2-FCV-70-87-B, As-Left, 7/18/2014.

k. WID: 117758166, MOVATS Testing of Motor Operated Valves, 1-FCV-070-0134-B, As-Left, 4/13/17.

l. WID: 119475524, VOTES Infinity Valve Diagnostic System Report, 2-FCV-070-0134-B, As-Left, 5/5/19.

m. WID: 118305123, MOVATS Testing of Motor Operated Valves, 1-FCV-070-0090-A, As-Left, 3/27/17.

n. WID: 112241713, MOVATS Testing of Motor Operated Valves, 2-FCV-070-0090-A, As-Left, 8/21/2013.

15. Watts Bar UFSAR Section 6.6, Containment Systems.

16. ANSI/ANS-56-8-1994, American National Standard for Containment System Leakage Testing Requirements.

17. TVA Engineering Work Request, EWR20MECH026076, Generate U2 Containment Isolation Valve List and Design Inputs for Kalsi Engineering Pa impact evaluation.

08/19/20.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 1 of 6 Wedge Gate Valve Leak Testing Input data from References 10 and 11:Value Description Variable Units 150-lb 900-lb Seat ring OD dOD 4.628 4.242 in Seat ring ID dID 3.975 3.469 in Mean seat ring diameter dm 4.3015 3.8555 in Stem diameter at packing ds 0.997 1.123 in Measured packing load Fpack-1 275 770 lb Measured Packing load Fpack-2 230 690 lb Measured Packing load Fpack-3 685 lb Measured disk-to-seat 0.25 Mu-1 0.136 friction (average)Measured disk-to-seat 0.13 Mu-2 0.158 friction (average)Measured disk-to-seat 0.115 Mu-3 friction (average)Half wedge angle 4.75 5 deg Area based on dm Ao 14.53 11.67 in^2 Area based on ds As 0.78 0.78 in^2 Mean seat diameter, dm, is the average of dOD and dID.Mechanical Seat Load, lb, is determined using Equation 6.( ) (6)

2 (sin + cos )Mechanical Seat Contact Load, lb/in, is determined by dividing Equation 6 by the seat contact circumference, *dm.DP Seat Load is determined using Equation 7_ = (7)DP Seat Contact Load, lb/in, is determined by dividing Equation 7 by the seat contact circumference,

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 2 of 6 4-inch, 150-lb Flexible Wedge Gate Valve, Reference 10 Disk Side A Pu Thrust Leakage Disk Closing Mech Seat Load DP Seat Load Total Seat Load psi lb ml Thrust lb lb lb/in lb lb/in lb lb/in Decrease Fpack = 275 lb, Mu = 0.136 50 4990 0 4676 10708 792 727 54 9981 739 0%100 5010 2 4657 10664 789 1453 108 9211 682 8%150 4980 3 4588 10506 777 2180 161 8326 616 17%200 5000 6 4569 10463 774 2906 215 7556 559 24%250 5000 9 4530 10373 768 3633 269 6740 499 32%300 5005 8 4496 10295 762 4360 323 5936 439 41%350 5000 31 4452 10195 754 5086 376 5108 378 49%365 4970 103 4410 10099 747 5304 393 4795 355 52%380 5000 350 4428 10141 750 5522 409 4619 342 54%400 5010 960 4423 10128 749 5813 430 4315 319 57%Fpack = 275 lb, Mu = 0.136 50 3985 2 3671 8406 622 727 54 7680 568 0%100 3970 16 3617 8283 613 1453 108 6830 505 11%150 4040 14 3648 8354 618 2180 161 6174 457 20%200 4005 290 3574 8184 606 2906 215 5278 391 31%230 3985 1250 3530 8085 598 3342 247 4742 351 38%260 4005 2075 3527 8077 598 3778 280 4298 318 44%Fpack = 275 lb, Mu = 0.136 150 5675 0 5283 12098 895 2180 161 9918 734 0%200 5665 0 5234 11986 887 2906 215 9079 672 8%250 5670 0 5200 11908 881 3633 269 8275 612 17%300 5705 3 5196 11898 880 4360 323 7539 558 24%350 5710 5 5162 11820 875 5086 376 6734 498 32%400 5725 12 5138 11765 871 5813 430 5952 440 40%450 5710 20 5084 11642 861 6539 484 5102 378 49%480 5735 78 5085 11645 862 6975 516 4670 346 53%500 5710 465 5045 11552 855 7266 538 4286 317 57%Fpack = 275 lb, Mu = 0.136 50 8005 0 7691 17612 1303 727 54 16886 1250 0%150 8035 0 7643 17502 1295 2180 161 15322 1134 9%250 8015 0 7545 17278 1279 3633 269 13645 1010 19%350 8040 0 7492 17156 1270 5086 376 12070 893 29%450 8055 2 7429 17012 1259 6539 484 10472 775 38%500 7980 1 7315 16751 1240 7266 538 9484 702 44%520 8035 1 7354 16841 1246 7557 559 9284 687 45%Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 3 of 6 Disk Side A Pu Thrust Leakage Disk Closing Mech Seat Load DP Seat Load Total Seat Load psi lb ml Thrust lb lb lb/in lb lb/in lb lb/in Decrease 550 8005 0 7301 16718 1237 7993 591 8726 646 48%Fpack = 275 lb, Mu = 0.136 150 6505 0 6113 13998 1036 2180 161 11819 875 0%250 6495 6 6025 13797 1021 3633 269 10164 752 14%350 6520 9 5972 13675 1012 5086 376 8589 636 27%400 6550 14 5963 13655 1010 5813 430 7842 580 34%450 6420 19 5794 13268 982 6539 484 6728 498 43%500 6500 26 5835 13361 989 7266 538 6095 451 48%550 6515 34 5811 13306 985 7993 591 5314 393 55%580 6525 35 5797 13276 982 8429 624 4847 359 59%Disk Side B Disk Closing Mech Seat Load DP Seat Load Total Seat Load Pu Thrust Leakage Thrust lb lb lb/in lb lb/in lb lb/in Decrease Fpack = 230 lb, Mu = 0.158 50 4000 1 3731 7764 575 727 54 7038 521 0%100 4105 21 3797 7902 585 1453 108 6448 477 8%120 3975 68 3651 7599 562 1744 129 5855 433 17%150 3980 500 3633 7560 559 2180 161 5380 398 24%160 4000 730 3645 7586 561 2325 172 5260 389 25%170 3995 860 3632 7559 559 2470 183 5088 377 28%180 3995 1043 3624 7543 558 2616 194 4927 365 30%200 3990 1460 3604 7500 555 2906 215 4593 340 35%Fpack = 230 lb, Mu = 0.158 50 5000 2 4731 9845 729 727 54 9119 675 0%100 5020 40 4712 9806 726 1453 108 8352 618 8%120 4980 130 4656 9690 717 1744 129 7946 588 13%130 4990 230 4659 9695 717 1889 140 7805 578 14%140 5025 410 4686 9751 722 2035 151 7717 571 15%150 5010 550 4663 9704 718 2180 161 7524 557 17%170 5005 1200 4642 9661 715 2470 183 7190 532 21%180 5005 1160 4634 9644 714 2616 194 7029 520 23%200 5000 1720 4614 9602 711 2906 215 6695 495 27%Fpack = 230 lb, Mu = 0.158 100 6470 0 6162 12823 949 1453 108 11370 841 0%150 6480 1 6133 12763 944 2180 161 10583 783 7%200 6475 3 6089 12671 938 2906 215 9765 723 14%250 6505 2 6080 12652 936 3633 269 9019 667 21%Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 4 of 6 Disk Side B Disk Closing Mech Seat Load DP Seat Load Total Seat Load Pu Thrust Leakage Thrust lb lb lb/in lb lb/in lb lb/in Decrease 300 6500 5 6036 12561 929 4360 323 8201 607 28%350 6495 9 5992 12469 923 5086 376 7383 546 35%370 6505 43 5986 12457 922 5377 398 7080 524 38%380 6505 66 5978 12441 921 5522 409 6919 512 39%400 6495 190 5953 12388 917 5813 430 6575 487 42%430 6505 790 5939 12360 915 6249 462 6111 452 46%Fpack = 230 lb, Mu = 0.158 200 6970 6 6584 13701 1014 2906 215 10795 799 0%300 6985 4 6521 13570 1004 4360 323 9210 682 15%400 7010 13 6468 13460 996 5813 430 7647 566 29%430 6995 52 6429 13380 990 6249 462 7131 528 34%450 6995 140 6414 13347 988 6539 484 6808 504 37%470 6995 400 6398 13315 985 6830 505 6484 480 40%490 6995 930 6382 13282 983 7121 527 6161 456 43%510 6995 1280 6367 13250 980 7411 548 5838 432 46%Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 5 of 6 4-inch, 900-lb Flexible Wedge Gate Valve, Reference 11 Disk Side A Pu Thrust Leakage Disk Closing Mech Seat Load DP Seat Load Total Seat Load psi lb ml Thrust lb lb lb/in lb lb/in lb lb/in Decrease Fpack = 770 lb, Mu = 0.25 100 4385 5 3516 5229 432 1167 96 4061 335 0%200 4405 18.5 3437 5111 422 2335 193 2776 229 32%300 4405 100 3338 4964 410 3502 289 1462 121 64%350 4405 275 3288 4890 404 4086 337 804 66 80%400 4405 405 3239 4817 398 4670 386 147 12 96%450 4405 430 3189 4743 392 5254 434 -511 -42 113%600 4405 560 3041 4522 373 7005 578 -2483 -205 161%Fpack = 690 lb, Mu = 0.25 100 7000 4 6211 9237 763 1167 96 8069 666 0%200 7050 11 6162 9164 757 2335 193 6829 564 15%300 7030 45 6043 8987 742 3502 289 5484 453 32%350 7055 190 6018 8950 739 4086 337 4864 402 40%400 6995 480 5909 8788 725 4670 386 4118 340 49%500 6950 490 5765 8573 708 5837 482 2736 226 66%600 6960 710 5676 8441 697 7005 578 1436 119 82%650 6955 820 5621 8360 690 7589 627 771 64 90%700 7010 700 5627 8368 691 8172 675 196 16 98%Fpack = 685 lb, Mu = 0.25 100 7045 1 6261 9311 769 1167 96 8144 672 0%200 7020 1 6137 9127 754 2335 193 6792 561 17%300 7000 0 6018 8950 739 3502 289 5447 450 33%400 6990 0 5909 8788 725 4670 386 4118 340 49%500 7015 1 5835 8677 716 5837 482 2840 234 65%700 7015 24 5637 8383 692 8172 675 210 17 97%800 7035 37 5558 8265 682 9340 771 -1075 -89 113%900 7015 290 5439 8088 668 10507 867 -2419 -200 130%1000 7020 860 5345 7948 656 11675 964 -3727 -308 146%1100 6995 620 5220 7764 641 12842 1060 -5079 -419 162%Fpack = 685 lb, Mu = 0.13 200 7160 0 6277 14486 1196 2335 193 12151 1003 0%400 7035 0 5954 13740 1134 4670 386 9070 749 25%500 7025 0 5845 13488 1114 5837 482 7651 632 37%700 7015 0 5637 13008 1074 8172 675 4836 399 60%900 7000 2 5424 12516 1033 10507 867 2009 166 83%1000 7000 1 5325 12288 1014 11675 964 613 51 95%Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 1 Appendix A Page 6 of 6 Disk Side A Pu Thrust Leakage Disk Closing Mech Seat Load DP Seat Load Total Seat Load psi lb ml Thrust lb lb lb/in lb lb/in lb lb/in Decrease 1100 6995 1 5220 12048 995 12842 1060 -795 -66 107%Fpack = 685 lb, Mu = 0.13 200 4415 0 3532 8151 673 2335 193 5816 480 0%400 4525 10 3444 7947 656 4670 386 3278 271 44%500 4490 20 3310 7638 631 5837 482 1801 149 69%600 4500 140 3221 7433 614 7005 578 428 35 93%700 4510 610 3132 7227 597 8172 675 -945 -78 116%800 4515 800 3038 7010 579 9340 771 -2330 -192 140%900 4525 530 2949 6805 562 10507 867 -3703 -306 164%Disk Side B Pu Thrust Leakage Disk Closing Mech Seat Load DP Seat Load Total Seat Load psi lb ml Thrust lb lb lb/in lb lb/in lb lb/in Decrease Fpack = 685 lb, Mu = 0.115 200 4505 0 3622 8978 741 2335 193 6643 548 0%400 4500 1 3419 8474 700 4670 386 3804 314 43%500 4560 3 3380 8377 692 5837 482 2540 210 62%600 4550 5 3271 8107 669 7005 578 1102 91 83%700 4545 22 3167 7849 648 8172 675 -323 -27 105%800 4545 290 3068 7604 628 9340 771 -1736 -143 126%900 4545 790 2969 7358 607 10507 867 -3149 -260 147%1000 4540 980 2865 7100 586 11675 964 -4575 -378 169%1100 4540 560 2765 6855 566 12842 1060 -5988 -494 190%Fpack = 685 lb, Mu = 0.115 200 7045 0 6162 15274 1261 2335 193 12939 1068 0%400 7000 0 5919 14671 1211 4670 386 10001 826 23%600 6990 0 5711 14155 1169 7005 578 7150 590 45%800 7055 1 5578 13825 1141 9340 771 4485 370 65%1000 7050 2 5375 13322 1100 11675 964 1647 136 87%1100 7055 2 5280 13089 1081 12842 1060 246 20 98%1200 7050 3 5176 12831 1059 14010 1157 -1179 -97 109%Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 2 Revisions Rev. DCR/N Pages No. No. Description of Changes Affected 0 N/A Initial release All Description Pages Main Text 19 Total Pages 19 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 3 Table of Contents Page 1 OBJECTIVE AND SCOPE 5 1.1 Objective 5 1.2 Scope 5 1.3 Historical Leakage 8 2 METHODOLOGY 9 2.1 Variables 9 2.2 plug Valve Sealing Load 9 2.3 Sleeve sealing torque due to actuator load 10 2.4 Sleeve Sealing torque due to Differential Pressure 10 2.5 Total Sealing Force and Seal Force Reduction 11 3 INPUTS 12 3.1 Calculation Inputs 12 3.1.1 Adjusted Maximum Containment Design Pressure 14 4 ASSUMPTIONS 15 5 RESULTS 16 5.1 SEAT torque Reduction 16 5.2 Maximum LLRT Test Pressure for 10% Seal torque Reduction 16 5.3 Notes/ Recommendations 16 6 REFERENCES 18 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 4 List of Tables Table Description Page Table 1-1: Analysis Scope 5 Table 1-2: LLRT Leakage History 8 Table 3-1: Input Data 13 Table 5-1: Seat Load Reduction 17 List of Figures Table Description Page Figure 1-1: A Cut-section of the Xomox Tufline Plug Valve with Socket Weld Ends and the Sleeve [11] 7 Figure 2-1: A schematic showing plug-to-sleeve reaction under zero DP (left) and positive DP (right) 11 Figure 3-1: Dimensions needed for the calculations 12 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 5 1OBJECTIVE AND SCOPE 1.1 OBJECTIVE Kalsi Engineering, Inc. (KEI) has been contracted by Tennessee Valley Authority (TVA) to provide engineering services to evaluate the impact of local leak rate test (LLRT) pressures (DPtest) greater than the calculated peak containment internal pressure (Pa) caused by the design-basis loss-of-coolant accident (LOCA) for cases where greater test pressure tends to increase the sealing force. This work is being done in accordance with Purchase Order No. 6232543.The objective of this report is to determine the impact of the reduced LLRT pressure from DPtest to Pa on the seat leakage. All work performed under this project is done in accordance with the requirements of the KEI Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B requirements.1.2 SCOPE The scope of this attachment is (Motor-Operated Valve) MOV and (Air-Operated Valve) AOV plug valves listed in Table 1-1. The Group 67-5 valves are MOVs whereas the Group 31-2 and 77-2 valves are AOVs. These valves are two-way 2-inch Xomox Tufline plug valves with socket weld ends [4]. A cut section of a valve is shown in Figure 1-1.Table 1-1: Analysis Scope Group Component Id Comp Description Manufacturer 31-2 2-FCV-31-305 INCORE INSTR RM AHU 2A CWR Xomox - Crane ISOL 31-2 2-FCV-31-306 INCORE INSTR RM AHU 2A CWR Xomox - Crane ISOL 31-2 2-FCV-31-308 INCORE INSTR RM AHU 2A CWS Xomox - Crane ISOL 31-2 2-FCV-31-309 INCORE INSTR RM AHU 2A CWS Xomox - Crane ISOL 31-2 2-FCV-31-326 INCORE INSTR RM AHU 2B CWR Xomox - Crane ISOL 31-2 2-FCV-31-327 INCORE INSTR RM AHU 2B CWR Xomox - Crane ISOL Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 6 Group Component Id Comp Description Manufacturer 31-2 2-FCV-31-329 INCORE INSTR RM AHU 2B CWS Xomox - Crane ISOL 31-2 2-FCV-31-330 INCORE INSTR RM AHU 2B CWS Xomox - Crane ISOL 67-5 UPPER CNTMT VENT CLR 1(2)A 1(2)-FCV-67-130 Xomox - Crane (67-6) ERCW SUP HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)A 1(2)-FCV-67-131 Xomox - Crane (67-6) ERCW RET HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)C 1(2)-FCV-67-133 Xomox - Crane (67-6) ERCW SUP HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)C 1(2)-FCV-67-134 Xomox - Crane (67-6) ERCW RET HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)B 1(2)-FCV-67-138 Xomox - Crane (67-6) ERCW SUP HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)B 1(2)-FCV-67-139 Xomox - Crane (67-6) ERCW RET HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)D 1(2)-FCV-67-141 Xomox - Crane (67-6) ERCW SUP HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)D 1(2)-FCV-67-142 Xomox - Crane (67-6) ERCW RET HDR ISOL 67-5 UPPER CNTMT VENT CLR 1(2)A 1(2)-FCV-67-295 Xomox - Crane (67-6) ERCW RET ISOL 67-5 UPPER CNTMT VENT CLR 1(2)C 1(2)-FCV-67-296 Xomox - Crane (67-6) ERCW RET ISOL 67-5 UPPER CNTMT VENT CLR 1(2)B 1(2)-FCV-67-297 Xomox - Crane (67-6) ERCW RET ISOL 67-5 UPPER CNTMT VENT CLR 1(2)D 1(2)-FCV-67-298 Xomox - Crane (67-6) ERCW RET ISOL RB SUMP DISCHARGE FLOW 77-2 1(2)-FCV-77-127 Xomox - Crane CONTROL RB SUMP DISCHARGE FLOW 77-2 1(2)-FCV-77-128 Xomox - Crane CONTROL Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 7 Stem seal Trapezoidal openings in the sleeve & the plug Figure 1-1: A Cut-section of the Xomox Tufline Plug Valve with Socket Weld Ends and the Sleeve [11]Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 8 1.3 HISTORICAL LEAKAGE References [3] and [13] provide the LLRT history for the valves. Table 1-2, below, summarizes the results.Table 1-2: LLRT Leakage History Group Component Id Leakage Results 31-2 2-FCV-31-305 Favorable history 31-2 2-FCV-31-306 Favorable history 31-2 2-FCV-31-308 Favorable history 31-2 2-FCV-31-309 Favorable history 31-2 2-FCV-31-326 Favorable history 31-2 2-FCV-31-327 Unfavorable history 31-2 2-FCV-31-329 Unfavorable history 31-2 2-FCV-31-330 Favorable history 67-5 1-FCV-67-130 Favorable history 67-6 2-FCV-67-130 Unfavorable history 67-5 1-FCV-67-131 Favorable history 67-6 2-FCV-67-131 Favorable history 67-5 1-FCV-67-133 Favorable history 67-6 2-FCV-67-133 Favorable history 67-5 1-FCV-67-134 Favorable history 67-6 2-FCV-67-134 Favorable history 67-5 1-FCV-67-138 Favorable history 67-6 2-FCV-67-138 Favorable history 67-5 1-FCV-67-139 Favorable history 67-6 2-FCV-67-139 Favorable history 67-5 1-FCV-67-141 Favorable history 67-6 2-FCV-67-141 Favorable history 67-5 1-FCV-67-142 Favorable history 67-6 2-FCV-67-142 Favorable history 67-5 1-FCV-67-295 Favorable history 67-6 2-FCV-67-295 Favorable history 67-5 1-FCV-67-296 Favorable history 67-6 2-FCV-67-296 Favorable history 67-5 1-FCV-67-297 Favorable history 67-6 2-FCV-67-297 Favorable history 67-5 1-FCV-67-298 Favorable history 67-6 2-FCV-67-298 Favorable history 77-2 1-FCV-77-127 Favorable history 77-2 2-FCV-77-127 Favorable history 77-2 1-FCV-77-128 Favorable history 77-2 2-FCV-77-128 Favorable history Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 9 2METHODOLOGY 2.1 VARIABLES Variable Description Units Dpo Port diameter Inch A Dimension A (see Figure 3-1) Inch B Dimension B (see Figure 3-1) Inch Ap Projected area over which the DP acts. Calculated as Dpo*(A+B)/2 Inch^2 R Average plug radius Inch

 µ Plug-to-sleeve friction coefficient -

Ptest Differential pressure used for LLRT Psi Pa Calculated peak containment internal pressure related to the design- Psig basis loss-of-coolant accident (LOCA)Unseating torque Ft*lbf

%Reduction Reduction in sealing load due to Pa vs. test DP %

2.2 PLUG VALVE SEALING LOAD During LLRT, the plug valve is closed and differential pressure (DP) is applied across the valve.In the closed condition, the seat torque on the plug is contributed by 1) actuator torque and 2) DP-induced torque. Actuator torque is proportional to the applied torque on top of the plug stem and coefficient of friction (COF) at the contacting surfaces. The plug contacts the sleeve and the stem seal shown in Figure 1-1. DP load is proportional to the DP and the area over which the DP acts.The approach for this analysis is to:

1. Calculate the seat torque in the closed condition of the valve due to actuator torque

2. Calculate the DP-induced torque on the seat

3. Determine the reduction in total sealing torque due to changing the LLRT test DP from the current value, Ptest, to the design basis containment pressure, Pa.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 10 2.3 SLEEVE SEALING TORQUE DUE TO ACTUATOR LOAD The plug valve in the assembled condition develops contact reactions at the sleeve and the stem seal. The actuator torque needs to overcome frictional resistance at these interfaces.Since the actuator torque used in the LLRT is not available, it was estimated using the available unseating torque measurements. It was assumed that 75% of the unseating torque is contributed by the plug-to-sleeve friction and the remaining 25% is contributed by the plug-to-stem seal friction. As will be seen later in the report, this assumption does not affect the final conclusion.Therefore, the seat torque can be calculated using Equation 1.

 = = 0.25 = 0.75 (1)

Figure 2-1 schematically shows the effect of zero DP and a positive DP on the valve under closed condition. Due to symmetry of the plug, sleeve and the valve body about the centerline shown in Figure 2-1, the plug-to-sleeve reaction would be symmetric about the centerline when there is no DP load acting on the plug. Therefore, the seat torque can be written as the sum of seat torque contributions from the upstream reaction ( ) and downstream reaction ( ) of the seat.__ = µ + µ = µ( + ) (2) where, µ is plug-to-sleeve coefficient of friction (COF) and R is the average plug radius.2.4 SLEEVE SEALING TORQUE DUE TO DIFFERENTIAL PRESSURE Differential pressure acts normal to the conical plug surface and results in a force equal to the differential pressure (DP) times area (Ap) over which the DP acts. This area of the trapezoidal opening in the sleeve shown in Figure 1-1. The DP force can be calculated as:

 = (3)

Due to symmetry of the plug, sleeve and the valve body about the centerline shown in Figure 2-1, under the action of a higher pressure on the upstream side, the plug-to-sleeve reaction would reduce on the upstream side and increase on the downstream side by the same amount. For the low DP LLRT conditions, the upstream seat compression preload would not completely relax. The other variable contributing to the seat torque is the plug-to-sleeve COF, which is not expected to change for the LLRT DP range of 9 to 16.5 psid. Therefore, the net effect of the DP load on the valve torque would be zero as explained in the below equation._ = µ_ + µ_ (4)

 = µ(_ + _ ) = µ( + + ) = µ( + ) = __

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 11 Therefore, the DP load will have no effect on the seat torque. This conclusion is in agreement with the valve manufacturers valve torque calculation report [5] in which its stated that, Differential pressure applied across the valve has negligible effect on the valve torque. Increased differential pressure acts to increase frictional forces on the downstream side of the seal; however, decrease such forces on the upstream side.Pressure Upstream Downstream Figure 2-1: A schematic showing plug-to-sleeve reaction under zero DP (left) and positive DP (right)Under the action of zero DP, the plug-to-sleeve reaction is symmetric about the centerline. Under the action of a higher pressure on the upstream side, the plug-to-sleeve reaction would reduce on the upstream side and increase on the downstream side by same amounts (the length of arrows indicate the force magnitude). The plug-to-sleeve COF is not expected to change for the LLRT pressure range of 9 to 16.5 psi. Therefore, the net effect on the valve torque would be negligible.2.5 TOTAL SEALING FORCE AND SEAL FORCE REDUCTION Based on the discussion in Section 2.4, the seat torque is independent of the DP load; especially for the range of DPs involved in the LLRT DP range of 9 to 16.5 psid. Therefore, reducing the LLRT pressure from 16.5 psid to 9 psid is expected to have negligible effect on the leakage.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 12 3INPUTS 3.1 CALCULATION INPUTS The dimensions needed for the calculations are shown in Figure 3-1. The required dimensions for the plug valves were not available from TVA. The port diameter was used from Reference [6].The other dimensions were obtained from a 2 Xomox plug valve with flanged ends procured by KEI for another project. These dimensions were used as the best available information. The unseating torque values were used from the MOVATs test reports for the individual valves.The input data for the analyses are documented in Table 3-1. Justified assumptions were made where data were not available. It should be noted that the results of this analysis will not be affected by changing the inputs.Figure 3-1: Dimensions needed for the calculations Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 13 Table 3-1: Input Data Item Variable Value Reference Port diameter, Inch Dpo 2.041 [6]Dimension A (see Figure 3-1), Inch A 2.252 Measured Dimension B (see Figure 3-1), Inch B 2.0975 Measured Projected area over which the DP acts, in^2 Ap 4.44 Calculated as Dpo*(A+B)/2 Average plug radius, inch R 1.087 Calculated as (A+B)/4 Plug-to-sleeve friction coefficient µ 0.23 [7]LLRT test differential pressure, psi DPtest 16.5 See 3.1.1 Adj. maximum containment design pressure, Pa 9.0 See 3.1.1 psig Unseating torque for 1-FCV-67-130, ft*lbf 2.3Note 1 [8]Unseating torque for 1-FCV-67-131, ft*lbf 3.1 Note 1 [8]Unseating torque for 1-FCV-67-133, ft*lbf 1.3 Note 1 [8]Unseating torque for 1-FCV-67-134, ft*lbf 2.4 Note 1 [8]Unseating torque for 1-FCV-67-138, ft*lbf 2.3 Note 1 [8]Unseating torque for 1-FCV-67-139, ft*lbf 1.5 Note 1 [8]Unseating torque for 1-FCV-67-141, ft*lbf 2.7 Note 1 [8]Unseating torque for 1-FCV-67-142, ft*lbf 2.7 Note 1[8]Unseating torque for 1-FCV-67-295, ft*lbf 3 Note 1 [8]Unseating torque for 1-FCV-67-296, ft*lbf 2.3 Note 1 [8]Unseating torque for 1-FCV-67-297, ft*lbf 2.1 Note 1 [8]Unseating torque for 1-FCV-67-298, ft*lbf 6 Note 1 [8]Actuator Overall Gear Ratio OAR OAR 47.85 [12]Actuator Pullout Efficiency act 0.35 [12]HBC Gear Ratio GR 70 [12]HBC Gear Efficiency HBC 0.3 [12]Note 1: The torque values were obtained using the MCC (Motor Control Center) method that measures actuator motor torque. The torque at the valve shaft can be obtained as =act . Based on the values of the OAR, act, GR & HBC, the multiplication factor to obtain the torque at valve shaft is 47.85*0.35*70*0.3 = 351.7. Therefore, the motor torque values should be multiplied by 351.7 to obtain the valve shaft torque.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 14 3.1.1 Adjusted Maximum Containment Design Pressure The maximum permissible LLRT test pressure is 1.1 x 15 = 16.5 psi [9][10]. The calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA) for Watts Bar is 9.36 psig [9]. A more conservative value of 9 psig is used in the current calculations.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 15 4ASSUMPTIONS Data that have not been formally verified are treated as assumptions. Where possible, the basis of the data has been noted. The following general assumptions were used in this analysis.

1. Reference [4] provided by TVA does not list all the plug valve tag IDs analyzed in this report. It is assumed that Reference [4] is applicable to valve tag IDs analyzed in this report.

This assumption do not affect the final conclusion and does not need verification.

2. 75% of the unseating torque is assumed to be contributed by the plug-to-sleeve friction and the remaining 25% is assumed to be contributed by plug-to-stem seal friction. This%

split between the two locations is reasonable and does not affect the final conclusion. This assumption does not need verification.

3. The plug has a small taper angle and therefore the DP force will have a small component along the plug axis. Based on the measurements of the 2 plug, the taper angle is approximately 2.75°. Therefore, the axial force component will be sin(2.75°) = 4.8% of the total load. This small axial load is neglected in the current analysis. This assumption does not need verification

4. The plug and sleeve material is stainless steel and UHMWPE. A COF of 0.23 for this material pair for LLRT test medium of air is used per Reference [7]. This is a reasonable assumption and does not affect the final conclusion. This assumption does not need verification.

5. Plug/port dimensions were used from a 2 welded-end Xomox valve procured by KEI for another project. These dimensions do not affect the final conclusion. This assumption does not need verification.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 16 5RESULTS 5.1 SEAT TORQUE REDUCTION Seating torque reduction for MOV plug valves due to a decrease in LLRT test pressure from the normal test DPtest (16 psi) to Pa (9 psi) is shown in Table 5-1. As can be seen, all values show 0%reduction in seat torque. Even though the data to calculate the seat load reduction for the AOV plug valves is not available, based on the discussion presented in Sections 2.4 and 2.5, the seat torques for all the plug valves would be independent of the DP load.5.2 MAXIMUM LLRT TEST PRESSURE FOR 10% SEAL TORQUE REDUCTION Since the seat torque is independent of the DP load, the maximum LLRT test pressure (DPtest) to ensure no greater than a 10% seal load reduction at Pa is not applicable to these valves.5.3 NOTES/ RECOMMENDATIONS None.Non-Proprietary Version

Non-Proprietary Version Document 3960C, Rev. 0, Attachment 2 Table 5-1: Seat Load Reduction Increase in Decrease in Net reduction in % reduction in seat Projected area Port diameter, Dimension A, Dimension B, Motor Unseating Avg. plug seat torque, Force due to friction torque on friction torque seat torque due to torque due to Group Valve Valve IDs over wh ich the DP In ch Inch In ch torque, ft* lbf torque, ft* lbf radius, in ft* lbf LLRT DP, lb! downstream seat, on upstream reduced LLRT reduced LLRT acts, inA2 ft* lbf seat, ft* lbf pressure, ft* lbf pressure,%Dp A B T motor Tunuat R Ap Tsaat Fo, 1-FCV 130 2.041 2.252 2.0975 2.3 808.90 1.09 4.44 606.68 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-131 2.041 2.252 2.0975 3.1 1090.26 1.09 4 .44 817.70 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-133 2.041 2.252 2.0975 1.3 457.21 1.09 4.44 342.91 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-134 2.041 2.252 2.0975 2.4 844.07 1.09 4 .44 633.06 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-138 2.041 2.252 2.0975 2.3 808.90 1.09 4 .44 606.68 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-139 2.041 2.252 2.0975 1.5 527.55 1.09 4 .44 395.66 39.95 0.83 -0.83 0 .00 0.00%67-5 plug valve - MOV 1-FCV-67-141 2.041 2.252 2.0975 2.7 949.58 1.09 4 .44 712.19 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-142 2.041 2.252 2.0975 2.7 949.58 1.09 4 .44 712.19 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-295 2.041 2.252 2.0975 3 1055.09 1.09 4 .44 791.32 39.95 0.83 -0.83 0 .00 0.00%1-FCV-67-296 2.041 2.252 2.0975 2.3 808.90 1.09 4.44 606.68 39.95 0 .83 -0.83 0 .00 0.00%1-FCV-67-297 2.041 2.252 2.0975 2.1 738.56 1.09 4.44 553.92 39.95 0.83 -0.83 0 .00 0.00%l-FCV-67-298 2.041 2.252 2.0975 6 2110.19 1.09 4 .44 1582.64 39.95 0 .83 -0.83 0 .00 0.00%I~6'illl m- Page 17 ci**

Document 3960C, Rev. 0, Attachment 2 Page 18 6REFERENCES[1] KEI Document No. 1500C Rev. 15; Kalsi Engineering, Inc. Quality Assurance Manual.[2] TVA Purchase Order 6232543, Rev. Num: 0.[3] TVA Engineering Work Request, EWR20MECH088032, Generate List of U1Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 06/09/20.[4] Figure 1366SW Size 2 Limitorque SMB-000-2-H1BC, Tufline Division of Xomox Corporation Spec No. NP3491-C Rev.902, 2-5-79[5] Calculation of Valve Torque for Tennessee Valley Authority Watts Bar Nuclear Plant, Xomox Corporation, Aug 1, 1989. Record No. B26 0929 945, TVA Ref. 89NNU-75552A-01, Xomox Ref. S. O. 90844[6] KEI Document No. 3738 Rev.0, ASME OM Code Appendix III Suitability Evaluation for Essential Raw Cooling Water MOVs 2-FCV-067-0130-A/0131-B/0133-A/0134-B/0138-B/0139-A/0141-B/0142-A/0295-A/0296-A/0297-B/0298-B, March 2018[7] EPRI Technical Report Technical Report 3002010634, Bearing Friction Coefficients for Quarter-Turn Valves Review of Available Information, September 2017[8] WBN Units 0, 1, & 2, Datasheets for MOVATs Testing Of Motor Operated Valves for MOVs: 1-FCV-67-130, 1-FCV-67-131, 1-FCV-67-133, 1-FCV-67-134, 1-FCV-67-138, 1-FCV-67-139, 1-FCV-67-141, 1-FCV-67-142, 1-FCV-67-295, 1-FCV-67-296, 1-FCV-67-297, 1-FCV-67-298.[9] Watts Bar UFSAR Section 6.2, Containment Systems.[10] ANSI/ANS-56.8-1994, American National Standard for Containment System Leakage Testing Requirements.[11] Xomox Process Valves & Actuators, Publication PN329703 04/06 3M C&O, Tufline Sleeved Plug Valves.[12] TVA WBN Calc No. MD000206720100376 Rev. 3, Document of Design Basis Review, Required Thrust/Torque Calculations, and Valve and Actuator Capability Assessment for Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 2 Page 19 Valves 2-FCV-67-130, -131, -133, -134, -138, -139, -141, -142, -295, -296, -297, -298, July 2018.[13] TVA Engineering Work Request, EWR20MEC026076, Generate U2 Containment Isolation Valve List and Design Inputs for Kalsi Engineering Pa impact evaluation.08/19/20.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 2 Revisions Rev. DCR/N Pages No. No. Description of Changes Affected 0 N/A Initial release All Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 3 Table of Contents Page 1 OBJECTIVE AND SCOPE 5 1.1 Objective 5 1.2 Scope 5 1.3 Historical Leakage 6 2 METHODOLOGY, CALCULATIONS, AND RESULTS 7 2.1 Variables 7 2.2 LLRT Test Pressure and Calculated Peak Pressure 8 2.3 Methodology for Swing Check Valve Sealing Load Analysis 8 2.3.1 Methodology for Soft-Seated Swing Check Valve 8 2.3.2 Methodology for Metal-Seated Swing Check Valve 9 2.4 Soft-Seated Swing Check Valves 9 2.4.1 Group 67-2 (Unit 1 and Unit 2) and Group 70-1 (Unit 1) Valves 10 2.4.1.1 Design Inputs for Group 67-2 (Unit 1 and Unit 2) and Group 70-1 (Unit 1) Valves 11 2.4.1.2 Peak Seat Contact Stress Calculation 11 2.4.2 Group 70-1 (Unit 2) Valve 13 2.4.2.1 Design Inputs for Group 70-1 (Unit 2) Valves 14 2.4.2.2 Peak Seat Contact Stress Calculation 14 2.5 Metal-Seated Swing Check Valves 16 2.5.1 Design Inputs for Groups 26-1 and 81-1 Valves 18 2.5.2 Assumptions for Groups 26-1 and 81-1 Valves 18 2.5.3 Calculation of Percentage Increase in Leakage Flow Area 18 3 ASSUMPTIONS 21 4 CONCLUSIONS AND RECOMMENDATIONS 23 4.1 Soft-Seated Swing Check Valves Analysis Conclusions 23 4.2 Metal-Seated Swing Check Valves Analysis Conclusions 23 4.3 Recommendations 24 5 REFERENCES 25 Appendix A - Supporting Documents Pages Rev Main Text 26 0 Appendix A 44 0 Total 70 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 4 List of Tables Table Description Page Table 1-1: Analysis Scope 5 Table 1-2: Unit 1 and Unit 2 LLRT Leakage History 6 Table 2-1: Peak Seat Contact Stress Calculation Results at O-Ring Shore A Hardness of 60 Durometer 12 Table 2-2: Percentage Increase in Leakage Flow Area Calculation Results 20 List of Figures Table Description Page Figure 2-1: Group 67-2 Unit 1 and Unit 2 Valve [9.b] 10 Figure 2-2: Group 70-1 Unit 1 Valve [9.c] 10 Figure 2-3: Compression Load per Linear Inch of Seal for 0.210 Cross Section O-ring and 60 Shore A Hardness [5] 13 Figure 2-4: Group 70-1 Unit 2 Valve [9.d] 14 Figure 2-5: Group 26-1 Unit 1 and Unit 2 Valves [9.a] 16 Figure 2-6: Group 81-1 Unit 1 and Unit 2 Valves [9.e] 16 Figure 2-7: Microscopic Flow Path Under Light and Heavy Seating Load [6] 17 Figure 2-8: Surface Asperities (High Spots) on a Seat Contact Band 18 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 5 1OBJECTIVE AND SCOPE 1.1 OBJECTIVE Kalsi Engineering, Inc. (KEI) has been contracted by Tennessee Valley Authority (TVA) to provide engineering services to evaluate the impact of local leak rate test (LLRT) pressures, DPtest, higher than the calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA), Pa, for cases where higher test pressure tends to increase the sealing.This work is being done in accordance with the scope defined in Purchase Order No. 6232543 [2]1.The objective of this report is to determine the impact of the reduced LLRT pressure from DPtest to Pa on the seat leakage. All work performed under this project was done in accordance with the requirements of the KEI Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B requirements.1.2 SCOPE The scope of this attachment is LLRT swing check valves. Component IDs and basic information are shown below in Table 1-1.Table 1-1: Analysis Scope Manufacturer/Group Component ID Comp Description Drawing No./ Seat Type 1-CKV-26-1260 REACTOR BLDG HPFP SUPPLY HDR BORG-WARNER CORP./2-CKV-26-1260 CHECK 421JBB1-002/26-1 1-CKV-26-1296 REACTOR COOLANT PUMP Hard Seat 2-CKV-26-1296 SPRINKLER HDR ISOL CHK 1-CKV-67-580A 1-CKV-67-580B A585-ATWOOD &1-CKV-67-580C UPPER CNTMT VENT CLR 1A, 1B, MORRILL CO/67-2 1C, 1D, 2A, 2B, 2C, 2D ERCW SUP 1-CKV-67-580D 14735-02/HDR CHECK 2-CKV-67-580A Soft Seat 2-CKV-67-580B 1The number in [] indicates reference number documented in Section 5.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 6 Manufacturer/Group Component ID Comp Description Drawing No./ Seat Type 2-CKV-67-580C 2-CKV-67-580D ATWOOD & MORRILL/1-CKV-70-679 14735-01/RCP THERMAL BARRIER CCS SUP Soft Seat 70-1 HDR CHECK FLOWSERVE/2-CKV-70-679 13-103681-001/Soft Seat W120-WESTINGHOUSE 1-CKV-81-502 PRIMARY WATER CNTMT HDR ELEC CORP/81-1 CHECK VLV 934D174/2-CKV-81-502 Hard Seat 1.3 HISTORICAL LEAKAGE Reference 3 and Reference 7 provide the LLRT history for the Unit 1 and Unit 2 AOVs. Table 1-2, below, summarizes the results.Table 1-2: Unit 1 and Unit 2 LLRT Leakage History Group Component ID Leakage Results Seat Type Notes 1-CKV-26-1260 Unfavorable history Hard None 1-CKV-26-1296 Unfavorable history Hard None 26-1 2-CKV-26-1260 Unfavorable history Hard None 2-CKV-26-1296 Favorable history Hard None 1-CKV-67-580A Favorable history Soft None 1-CKV-67-580B Favorable history Soft None 1-CKV-67-580C Favorable history Soft None 1-CKV-67-580D Favorable history Soft None 67-2 2-CKV-67-580A Favorable history Soft None 2-CKV-67-580B Favorable history Soft None 2-CKV-67-580C Favorable history Soft None 2-CKV-67-580D Favorable history Soft None 1-CKV-70-679 Unfavorable history Soft None 70-1 2-CKV-70-679 Favorable history Soft None 1-CKV-81-502 Favorable history Hard None 81-1 2-CKV-81-502 Favorable history Hard None Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 7 2METHODOLOGY, CALCULATIONS, AND RESULTS 2.1 VARIABLES Variable Description Units Am Area based on mean seat diameter = In2 AID Area based on seal ID = In2 Ahs Cross-sectional area of surface asperities In2 As Nominal seat contact area of swing check valve = In2 Percentage increase in leakage flow area when pressure reduces from AI %DPtest to Pa Percentage area of the nominal seat contact area due to asperity Aasp %contacts (actual seat contact area) b O-ring contact area per linear inch of seal In2/in CL Seat leakage coefficient proportionality constant -d O-ring cross-section diameter In da Diameter of surface asperities In DID Seal inside diameter In Dm Mean seat diameter In DOD Seal outside diameter In DP Valve differential pressure Psi DPtest Bounding LLRT test differential pressure Psi E Youngs modulus Psi F O-ring compression load Lb FDP Seat load due to differential pressure Lb FDP_DPtest Seat load due to differential pressure @ DPtest Lb FDP_DPtest_Lin Seat load per linear inch of seat diameter @ DPtest Lb/in FDP_Pa Seat load due to differential pressure @ Pa Lb FDP_Pa_Lin Seat load per linear inch of seat diameter @ Pa Lb/in Fhs Seat load per asperity (high spot) Lb FW Sealing force due to disc and stem weight Lb f Peak seat contact stress Psi favg Average seat contact stress Psi ha Height of the asperities (high spots) at no pressure/seat load In Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 8 ha_DPtest Height of the asperities (high spots) at DPtest In ha_Pa Height of the asperities (high spots) at Pa In ha Change in the height of asperities (high spots) In Nhs Number of high spots on the seat contact band Calculated peak containment internal pressure related to the design-Pa Psig basis loss-of-coolant accident (LOCA) t Seat contact band width of swing check valve In W Disc and stem weight Lb x O-ring deflection for a given% O-ring compression in Percentage reduction in sealing force due to the difference between R %Pa and DPtest 2.2 LLRT TEST PRESSURE AND CALCULATED PEAK PRESSURE The maximum permissible LLRT test pressure, DPtest, is 1.1 x 15= 16.5 psig [8, 12]. Calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA),Pa, for Watts Bar is 9.36 psig [8]. For purposes of this analysis, a lower and more conservative value of 9 psig is used.2.3 METHODOLOGY FOR SWING CHECK VALVE SEALING LOAD ANALYSIS The sealing load in a swing check valve, without any mechanical load, is equal to the differential pressure force acting on the disc2. The DP force, FDP, is equal to the DP times the area over which the DP acts.For the subject swing check valves, the percentage reduction in the sealing load, R, when the differential pressure reduces from DPtest to Pa is equal to the percentage reduction in the pressure which is given by:16.5 9.0

 = 100 = 100 = 45.5% (1) 16.5 This assessment includes soft-seated and metal-seated swing check valves.

2.3.1 Methodology for Soft-Seated Swing Check Valve In a soft-seated swing check valve, a low modulus elastomeric seal will drape into the surface asperities of the metal seat under low sealing forces generated at low DP conditions. The LLRT DP range of 9 to 16.5 psig represent low DP conditions. Therefore, the seat leak path that forms due to the surface asperity contacts will not be present in the soft-seated swing check valves. The 2The seat load component due to the disc and arm weight is very small compared to the DP force; therefore, it is excluded from the seat load which is conservative.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 9 following approach is used for the soft-seated valves to analyze the risk of increased leakage when the pressure is reduced to Pa:

a. Determine the peak seat contact stresses based on the seat load at DPtest and Pa pressures. The peak seat contact stress higher than the differential pressure will ensure a positive sealing margin at that pressure.

b. Determine the O-ring compression at DPtest and Pa pressures. The O-ring compression above the minimum recommended squeeze of 0.007 inches (0.2 mm) [5] will ensure a good sealing action.

2.3.2 Methodology for Metal-Seated Swing Check Valve Unlike the soft-seated swing check valves, a tight sealing of a metal-seated valve requires yielding of one material into the surface waviness and surface roughness of the other to block direct leakage paths. At low pressures, the seat load will not be sufficient to plastically yield the asperities on the contacting surfaces and therefore, the asperities will deform elastically. Due to the fact that these valves were providing a reliable sealing at DPtest pressure, the microscopic leak paths at this pressure will have a very high flow resistance. Any reduction in the seat load will decrease the leak path flow resistance by increasing the leakage flow area due to a reduced elastic deformation of the asperities (high spots). This increase in the leakage flow area results in corresponding increase in the leakage coefficient, CL. Based on Equation 3-3 in Section 3.2 of the main report, the CL can increase by as much as 35% before the measured leakage at pressure Pa would exceed the measured leakage at DPtest. The following approach is used for the metal-seated valves to analyze the risk of increased leakage when the pressure is reduced to Pa:

a. The reduction in the seat load due to decrease in pressure from DPtest to Pa will increase the leakage flow area due to a reduced elastic deformation of the asperities (high spots).

The increase in the leakage flow area will decrease the leakage flow resistance and result in a corresponding increase the leakage flow coefficient, CL. The percentage increase in the leakage flow area, AI, will be calculated to determine the corresponding effect on CL.

b. If the calculated change in the leakage flow area is found to be negligible, the leak path resistance and the leakage coefficient, CL, will remain constant. This will ensure that the increase in leakage coefficient, CL, will be well below 35%. Therefore, the measured leakage at the lower pressure, Pa, will not be higher than the measured leakage at the higher pressure, DPtest.

2.4 SOFT-SEATED SWING CHECK VALVES The valve Groups 67-2 and 70-1 have soft-seated disc (Table 1-1). The valves in Group 67-2 Units 1 and 2, and Group 70-1 Unit 1 has O-ring 2-332 [9.b, 9.c] on the disc to provide a soft sealing.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 10 The valve in Group 70-1 Unit 2 has a special design resilient seat [9.d] to provide a sealing at low pressures along with a Stellite-6 hardfaced metal seat to provide sealing at higher pressures.2.4.1 Group 67-2 (Unit 1 and Unit 2) and Group 70-1 (Unit 1) Valves The valves in Group 67-2 (Unit 1 and Unit 2) and Group 70-1 (Unit 1) are shown in Figure 2-1 and Figure 2-2. The sealing load, FDP, on these valves is calculated from the DP area, AID, and the differential pressure, DP.Figure 2-1: Group 67-2 Unit 1 and Unit 2 Valve [9.b]Figure 2-2: Group 70-1 Unit 1 Valve [9.c]Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 11 2.4.1.1 Design Inputs for Group 67-2 (Unit 1 and Unit 2) and Group 70-1 (Unit 1) Valves These valves has O-ring 2-332 [9.b, 9.c] providing a soft sealing. O-ring 2-332 has a nominal cross-section, d, nominal seal ID, DID, and nominal seal OD, DOD, of 0.210 inches, 2.350 inches, and 2.770 inches respectively [5]. The durometer hardness of the O-ring is not known but typically an O-ring with a Shore A Hardness of 70-durometer that has approximate room temperature Youngs modulus of 1040 psi [4] is used in such applications. A softer O-ring with a lower Youngs modulus will provide a conservative calculation of the peak seat contact stress; therefore, the O-ring Shore A Hardness of 60-durometer with Youngs modulus of 630 psi [4] is used in this analysis. The percentage O-ring compression will be determined at 70 durometer and 60 durometer Shore A hardness to ensure that the sufficient O-ring compression is achieved at pressure Pa.2.4.1.2 Peak Seat Contact Stress Calculation The seat load at DPtest and Pa are calculated below:_ = = 2.350 16.5 = 71.57 $%4 (2)_ = = 2.350 9.0 = 39.04 $%4 (3)The peak seat contact stress is calculated using Equation 4 [4].4

 &' = (4)

Where, F = O-ring compression load, lb b = O-ring contact area per linear inch of seal, in2/in x = O-ring deflection for a given% seal compression, in Dm = Mean seat diameter, in The O-ring compression force, F, depends on the O-ring deflection, x, and is calculated using Equation 5 [4],

 , ../ , 1 = ( ) *1.25 + - + 50 + - 2 (5)

( (To calculate the peak seat contact stress, f, the O-ring contact area per linear inch of the seal, b, is required which is calculated from the seal deflection, x, [4, 10].Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 12

 % = 2.4 , (6)

The O-ring deflection, x, is calculated by solving Equation 5 for x. The O-ring compression force, F, in Equation 5 is set equal to the seat load due to the DP force, FDP, which is calculated in Equations 2 and 3. The percentage O-ring compression is calculated from the seal deflection, x, using Equation 7.

 % 3 4567 89:;4<==596 = 100

((7)The average seat contact stress is calculated using Equation 8,

 & >? = (8)

Table 2-1 documents the peak seat contact stresses, percentage O-ring compression, and average seat contact stresses calculated at DPtest and Pa pressures.Table 2-1: Peak Seat Contact Stress Calculation Results at O-Ring Shore A Hardness of 60 Durometer LLRT Pressure DPtest Pa Pressure, psi 16.50 9.00 Seat Load due to DP (FDP), lb 71.57 39.04 Seat Load Per Linear Inch, lb/in 8.90 4.85 O-ring Compression (x), in 0.0298 0.0200 Percentage O-ring Compression,% 14.2 9.5 O-ring Contact Area Per Linear Inch (b), in2/in 0.072 0.048 Calculated O-ring Compression Load (F), lb 71.53 39.13 Peak Seat Contact Stress (f'), psi 158.34 129.06 Average Seat Contact Stress (f'avg), psi 123.53 101.36 As shown in Table 2-1, the calculated percentage O-ring compressions, at DPtest and Pa pressures, are 14.2% and 9.5% for Shore A Hardness of 60 durometer. The percentage O-ring compression will decrease at 70 durometer hardness. The percentage O-ring compression at pressure Pa for Shore A hardness of 70 durometer is 6.8% (0.014 inches) which is higher than the minimum Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 13 recommended squeeze of 0.007 inches (0.2 mm) [5]. The Parker O-ring Handbook [5] documents the compression load per linear inch of seal for 0.210 cross-section O-ring which is shown in Figure 2-3. Per Figure 2-3, the compression load per linear inch ranges from approximately 5 lb/in to 12 lb/in for Shore A Hardness of 60-durometer and at 10% compression. The seat load per linear inch for the subject O-ring at the pressures DPtest and Pa are 8.90 lb/in and 4.85 lb/in (Table 2-1) which is within the range discussed above. The calculated peak seat contact stresses are 158.34 psi at DPtest (16.5 psig) and 129.06 psi at Pa (9.0 psig). The calculated average seat contact stresses at DPtest and Pa are 123.53 psi and 101.36 psi, respectively.The percentage reduction in the sealing load, R, is 45.5% when the pressure is reduced from 16.5 psig to 9.0 psig. However, the peak and average seat contact stresses are well above the DPtest and Pa pressures. The peak contact stress well above the differential pressure will ensure a proper sealing. Therefore, the leakage is not expected to increase when the pressure is reduced from DPtest to Pa.Figure 2-3: Compression Load per Linear Inch of Seal for 0.210 Cross Section O-ring and 60 Shore A Hardness [5]2.4.2 Group 70-1 (Unit 2) Valve The valve in Group 70-1 (Unit 2) is shown in Figure 2-4. The sealing load, FDP, on this valve is calculated from the DP area, AID, and the differential pressure, DP.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 14 (a) (b)Figure 2-4: Group 70-1 Unit 2 Valve [9.d]2.4.2.1 Design Inputs for Group 70-1 (Unit 2) Valves This valve has a resilient elastomer seat (Item 306 in Figure 2-4b) to provide sealing at low pressures along with a Stellite-6 hardfaced metal seat (Item 004 in Figure 2-4b) to provide sealing at higher pressures. The mean seat diameter, Dm, of the seal is 3.170 inches [9.d]. KEI has performed Finite Element Analysis (FEA) on a similar design and size (3-inch) valve [11]. Based on Reference 11, the resilient seal seat ring groove ID, DID, is 2.938 inches, seal height/diameter, d, is 0.194+/-0.006 inches, and the seal projection above the seal groove (maximum seal deflection) is 0.025+/-0.011 inches. To be conservative, the seal height, d, of 0.188 inches, and maximum seal deflection, x, of 0.014 inches corresponding to seal least material condition are used in the analysis.This gives the maximum seal compression of 7.45% (0.014/0.188*100). The seal dimensions scaled from the valve drawing [9.d] matches with the dimensions in Reference 11. The durometer hardness of the seal is not known but typically a seal with a Shore A Hardness of 70-durometer that has approximate room temperature Youngs modulus of 1040 psi [4] is used in such applications. A softer seal with a lower Youngs modulus will provide a conservative peak seat contact stress; therefore, the seal Shore A Hardness of 60-durometer with Youngs modulus of 630 psi [4] is used in this analysis.2.4.2.2 Peak Seat Contact Stress Calculation The seat load at DPtest and Pa are calculated below using Equations 2 and 3:Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 15_ = = 2.938 16.5 = 111.86 $%4_ = = 2.938 9.0 = 61.01 $%4 Equations 4 and 5 are typically used to calculate the peak seat contact stresses and compression load for an O-ring. The same equations are used to estimate the peak seat contact stresses and compression load for the subject valve seal.The compression load, F, is calculated using Equation 5 and is given by:

 , ../ , 1 = ( ) *1.25 + - + 50 + - 2

( (

 . 014 ../ . 014 1 = 0.188 3.17 630 A1.25 B C + 50 B C D = 29.97 $% . 188 . 188 Equation 6 is used to calculate the seal contact area per linear inch, b, and is as follow: % = 2.4 , = 2.4 0.014 = .034 56 /56 The peak contact stress, f, is calculated using Equation 4 and is given by:

4 4 29.97

 &' = = = 114.05;=5 % 0.034 3.17 The maximum allowable seal compression, x, of the subject valve occurs at the compression load of 29.97 lb and develops the peak seat contact stresses of 114.05 psi which is higher than DPtest and Pa pressures. Based on FEA performed by KEI on a similar valve design [11], the peak seat contact stresses were found to be much higher even at 1 psi differential pressure. Therefore, based on an engineering judgement, it is concluded that the use of Equations 4 and 5 for calculating the peak seat contact stresses and compression load is conservative. The seal compression load of 29.97 lb is significantly lower than the DP induced seat load of 111.86 lb and 61.01 lb at DPtest and Pa pressures. This shows that the seal will stay fully compressed at DPtest and Pa and will develop peak seat contact stresses high enough to ensure good sealing action. Therefore, leakage is not expected to increase when the pressure is reduced from DPtest to Pa.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 16 2.5 METAL-SEATED SWING CHECK VALVES The valves in Groups 26-1 and 81-1 have metal-seated disc (Table 1-1) which is shown in Figure 2-5 and Figure 2-6.Figure 2-5: Group 26-1 Unit 1 and Unit 2 Valves [9.a]Figure 2-6: Group 81-1 Unit 1 and Unit 2 Valves [9.e]Unlike the soft-seated swing check valves, a tight sealing of a metal-seated valve requires yielding of one material into the surface waviness and surface roughness of the other to block direct leakage paths. Even, seemingly smooth machined surfaces have surface asperities as illustrated in Figure 2-7. When the two surfaces contact each other, the surface asperities initially establish the contact. With an increasing load, the asperities initially deform elastically and then plastically. To ensure a tight shut-off metal-to-metal seal, the surface asperities within the contact band need to deform plastically over a reasonable amount of bandwidth. At low pressures, the seat load will not be sufficient to plastically yield the asperities on the contacting surfaces and therefore, the Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 17 asperities will deform elastically. Due to the fact that these valves were providing a reliable sealing at DPtest pressure, the leak path at this pressure will have a very high flow resistance. Any reduction in the seat load will decrease the flow resistance by increasing the leakage flow area, AI, due to a reduced elastic deformation of the asperities (high spots). As discussed in Section 2.3, the seat load will decrease by 45.5% when the pressure reduces from the DPtest of 16.5 psig to Pa of 9.0 psig.The reduction in the seat load will reduce the leak path flow resistance. A reduction in the flow resistance is equivalent to an increase in the leakage coefficient, CL. Based on Equation 3-3 in Section 3.2 of the main report, the CL can increase by 35% before the measured leakage at pressure Pa would increase from the measured leakage at DPtest.Figure 2-7: Microscopic Flow Path Under Light and Heavy Seating Load [6]A calculation has been performed to estimate an increase in the leakage flow area, AI, when the pressure is reduced from DPtest to Pa. The calculation is based on a simple, but conservative assumption (see Section 2.5.2) of a leak flow area developed by surface asperities (high spots) on the disc/seat surfaces that comes in contact when the swing check valve disc closes. The idea behind this calculation is to determine the effect of the seat load change on the height of the asperities that in turn increases or decreases the leakage flow area. Figure 2-8 shows asperities of an exaggerated size on a seat contact band in the top view. For better visualization, only one high spot is shown in the side view. The side view shows a leakage flow channel developed by a high spot at zero seat load, at DPtest, and at Pa. The flow channel width is constant, whereas the height of the asperities, ha, changes as it gets compressed due to seat load.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 18 Figure 2-8: Surface Asperities (High Spots) on a Seat Contact Band For better visualization, only one asperity is shown in the side view. The side view shows the flow channel at no pressure, at DPtest, and at Pa. The flow channel width is constant, whereas the height, ha, changes as the seat load compresses the surface asperities.2.5.1 Design Inputs for Groups 26-1 and 81-1 Valves The design inputs used in the calculations are as follow:

The LLRT test pressure, DPtest, and calculated peak pressures, Pa, are documented in Section 2.2.
The mean seat diameter, Dm, of the valve Groups 26-1 and 81-1 are 4.625 inches and 3.590 inches [7] respectively.

2.5.2 Assumptions for Groups 26-1 and 81-1 Valves Assumptions 1 to 5 documented in Section 3 are used for the calculation of percentage increase in the leakage flow area.2.5.3 Calculation of Percentage Increase in Leakage Flow Area The percentage increase in leakage flow area, AI, is calculated using Equation 9. The AI equation does not account for a change in diameter of the asperities, da, due to a lateral strain while under Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 19 compression. The lateral strain of the asperities will change the effective width of the flow channel (see Figure 2-8). The change in effective width of the flow channel will be orders of magnitude smaller than the flow channel width and will have a negligible effect on AI.

 = x 100 (9)

Where,

_ = _ (10)_ = _ (11)The change in height of asperities due to the seat load is calculated using Equation 12:I

 = (12)

I )

Where, I = (13)

JI I = (4 (14)K 100 (15)JI =I Table 2-2 documents the percentage increase in the leakage flow area, AI.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 20 Table 2-2: Percentage Increase in Leakage Flow Area Calculation Results Valve Group 26-1 81-1 Dm, in 4.625 3.590 2As, in 1.45 1.13 FDP_DPtest, lb 277.20 167.02 FDP_Pa, lb 151.20 91.10 Nhs 7.23E+08 5.61E+08 Fhs_DPtest, lb 3.8E-07 3.0E-07 Fhs_Pa, lb 2.1E-07 1.6E-07 ha_DPtest, in 1.0E-03 7.9E-04 ha_Pa, in 5.6E-04 4.3E-04 AI,% 0.003 0.002 Table 2-2 shows that the calculated increase in the leakage flow area, AI, is 0.003% and 0.002%for Groups 26-1 and 81-1 valves which is negligible and is not expected to increase the leakage coefficient, CL, by 35% which is a threshold for the measured leakage at the lower pressure, Pa, to increase from the measured leakage at the higher pressure, DPtest.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 21 3ASSUMPTIONS Data that have not been formally verified are treated as assumptions. Where possible, the basis of the data has been noted. The following assumptions are used for the calculations performed in Section 2.5.3 for Groups 26-1 and 81-1 metal-seated valves:

1. The surface asperities (high spots) are assumed to be of a cylindrical profile. This is a simple geometry. Typically, the asperities have a profile with a smaller cross-sectional area at the contacting surfaces and the area increase towards the base of the asperities. Such profile will be much stiffer compared to a cylindrical profile. Therefore, the cylindrical profile will undergo a larger deformation due to the differential pressure force compared to the profile mentioned above. The larger deformation will result into a larger change in the leakage flow area for the which is conservative. Therefore, this assumption does not require a verification.

2. The height, ha, and diameter, da of the surface roughness asperities are assumed to be equal to 16 µin. This assumption is based on a surface roughness assumption of 16 Ra. Typically, the sealing surfaces are machined to the lowest surface roughness to ensure tight sealing.

A sensitivity analysis showed that this assumption has no effect on the calculated percentage increase in the leakage area. Therefore, this assumption does not require a verification.

3. The seat contact band width, t, is assumed to be 0.10 inches. Typically, for a swing check valve, the seat contact width is larger than 0.10 inches. A lower seat contact width provides conservative results. Therefore, this assumption does not require a verification.

4. It is assumed that the area covered with the surface asperities, Aasp, is 10% of the nominal seat contact area, As. Typically, the area covered with the asperities is higher than 10%. The seat load, FDP, acts on the area covered with the asperities therefore, lower the area, higher will be the deformation of the asperities which is conservative. A sensitivity analysis showed that lowering this value to 1% does not change the overall conclusion because the percentage increase in the leakage area remains negligible. Therefore, this assumption does not require a verification.

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Document 3960C, Rev. 0, Attachment 3 Page 22

5. The modulus of elasticity, E, of the seat/disc material is assumed to be equal to 3.0e7 psi.

The valve drawings [9.a and 9.e] shows that the disc and seat surfaces are hardfaced which typically has a higher modulus than the one used here. A lower modulus provides a conservative result. Therefore, this assumption does not require a verification.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 23 4CONCLUSIONS AND RECOMMENDATIONS The subject valve groups include soft-seated and metal-seated swing check valves.4.1 SOFT-SEATED SWING CHECK VALVES ANALYSIS CONCLUSIONS The soft seal of check valves provides very reliable sealing action at low differential pressure conditions. The LLRT DP range of 9 to 16.5 psig represent low DP conditions. Industry experience shows that in general, the soft-seated valves are expected to provide a good sealing action for this DP range. Calculations performed by KEI further solidify this general conclusion.The soft-seated disc of the valves in Groups 67-2 (Unit 1 and Unit 2) and 70-1 (Unit1) have an O-ring that provides a soft sealing. The percentage reduction in the sealing load, R, is 45.5% when the pressure is reduced from 16.5 psig to 9.0 psig (see Section 2.3). However, both, the peak and average seat contact stresses are well above the differential pressure at both DPtest and Pa pressures (Table 2-1). The peak seat contact stress must exceed the differential pressure to ensure sealing.Therefore, the leakage is not expected to increase when the pressure is reduced from DPtest to Pa.The valve in Group 70-1 Unit 2 has a special design resilient seat to provide a sealing at low pressures along with a Stellite-6 hardfaced metal seat to provide sealing at higher pressures. The calculation showed that maximum allowable seal compression of the subject valve occurs at the compression load of 29.97 lb which develops the peak seat contact stresses of 114.05 psi. The DP induced seat loads of 111.86 lb and 61.01 lb at DPtest and Pa pressures (see Section 2.4.2.2) are higher than the compression load (29.97 lb) required to fully compress the seal and establish a metal-to-metal contact. Therefore, at DPtest and Pa pressures, the seal will remain fully compressed and will establish a metal-to-metal sealing. The peak seat contact stresses at DPtest and Pa pressures will be significantly higher than the peak seat contact stress of 114.05 psi calculated at 29.97 lb of seat load. Therefore, the leakage is not expected to increase when the pressure is reduced from DPtest to Pa.4.2 METAL-SEATED SWING CHECK VALVES ANALYSIS CONCLUSIONS The valves in Groups 26-1 and 81-1 have metal-seated disc. The percentage reduction in the sealing load, R, for these valves is 45.5% when the pressure is reduced from 16.5 psig to 9.0 psig Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 24 (see Section 2.3). The reduction in the seat load will increase the leakage coefficient, CL, by some amount by increasing the leakage flow area due to a reduced elastic deformation of the asperities (high spots). The leakage coefficient, CL, is discussed in Section 3.2 of the main report which determines that the CL can increase by 35% before the measured leakage at the lower pressure, Pa, would increase from the measured leakage at the higher pressure, DPtest. Table 2-2 shows that the calculated percentage increase in the leakage flow area, AI, is negligible. The negligible increase in the leakage flow area will result in a negligible (well below 35%) increase in the leakage coefficient, CL. Therefore, the measured leakage at pressure Pa is not expected to increase from the measured leakage at pressure DPtest.4.3 RECOMMENDATIONS Based on the simplified but conservative calculation performed in Section 2.5 for the metal-seated swing check valves (Group 26-1 and 81-1), it is expected that the change in the leakage flow area will be negligible with the reduction in seat load. Therefore, it is expected that the leakage coefficient, CL, will not increase by 35% which is a threshold for the measured leakage at the lower pressure, Pa, to increase from the measured leakage at the higher pressure, DPtest. However, to further support the conclusion, it is recommended to test at least one metal-seated swing check valve from each group.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 25 5REFERENCES

1. KEI Document No. 1500C Rev. 15; Kalsi Engineering, Inc. Quality Assurance Manual.

2. TVA Purchase Order 6232543, Rev. Num: 0.

3. TVA Engineering Work Request, EWR20MEC088032, Generate List of U1 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 06/09/20.

4. Daniel L. Hertz, Machine Design, O-Ring for Low-Pressure Services, April 12, 1979.

5. ORD-5700, Parker O-Ring Handbook.

6. ISA Handbook of Control Valves, 2nd Edition, J.W. Hutchison, Instrument Society of America, 1979.

7. TVA Engineering Work Request, EWR20MEC026076, Work Order # 121532992, Generate U2 Containment Isolation Valve List and Design Inputs for Kalsi Engineering Pa Impact Evaluation, 08/19/20.

8. WBN UFSAR Section 6.2, Containment Systems.

9. Valve Drawing:

a. Borg-Warner Corp., Valve Assembly - 4 Inch, 150 LBS, Swing Check Valve, C.S.,

Dwg. No. 421JBB1-002, Rev. A.

b. Atwood & Morrill Co. Inc., 2 IN Class 210 WE Swing Check Valve W/Soft Seat Disc, Dwg. No. 14735-02, Approved Dated: 07-14-1981.

c. Atwood & Morrill Co. Inc., 3 IN Class 210 WE Swing Check Valve W/Soft Seat Disc, Dwg. No. 14735-01, Approved Dated: 11-20-1981.

d. Flowserve, Swing Check Valve Carbon Steel, Weld Ends with Resilient Seat Size:

3 Class: 150, Dwg. No. 13-103681-001, Rev. A, Dated: 08-01-13.

e. Westinghouse Electric Corporation, Swing Check Valve Model 03000CS8200000, 3.50 ASME CL. 1, GPO ASSY, Dwg. No. 934D174, Rev. A, Dated: 04-10-85.

10. HpA-O-ring Seal Design Best Practices, Rev. 1, Dated: 12-15-12.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 3 Page 26

11. KEI Document No. 2628, Rev. 2, Dual Seat Check Valve Soft Seat Analysis, Dated: 05 2009.

12. ANSI/ANS-56.8-1994, American National Standard for Containment System Leakage Testing Requirements.

Non-Proprietary Version

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 1 of 44 Appendix A SUPPORTING DOCUMENTS Page No.Title Page 1A Reference 4 2A Reference 5 9A Reference 6 25A Reference 9 33A Reference 10 38A Total Pages 44A Non-Proprietary Version

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 2 of 44 REPRINTED FROM MACHINE DESIGN April 12, 1979 O-RINGS FOR LOW-PRESSURE SERVICE P.O. BOX 519 , RED BANK, NEW JERSEY 07701 (201) 747-9200

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 3 of 44 DANIEL L. HERTZ, JR.President Seals Eastern Inc. Red Bank, N.J.O-RINGS normally operate with about 15% squeeze to ensure a tight seal.But at system pressures below 400 psi, this amount of squeeze can cause high friction and excessively high actuating forces.Reducing the amount of squeeze lowers friction to acceptable levels; however, lower squeeze also means lower sealing pressure and greater potential for leakage. This problem is aggravated by the stress relaxation characteristics of the seal material.Thus, an O-ring that seals well initially may lose resilience with time and fail suddenly.Designing O-ring seals for low pressures, therefore, is not simply a matter of reducing the amount of squeeze: it involves a delicate balancing of material hardness, dimensional tolerances, stress relaxation, and friction characteristics.Material Hardness The initial phase of designing a low-pressure O-ring seal is the same as that for a conventional O-ring.Size and fluid compatibility requirements are evaluated and O-ring dimensions selected from a catalog. The catalogs usually list a recommended range of squeeze values, as shown in Table 1.Squeeze is defined as the ratio 2

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 4 of 44 Most O-rings operate with enough "squeeze" to provide a reliable seal under almost any conditions. But low-pressure systems generate less squeeze, increasing the potential for leakage. Only a careful balancing of O-ring material properties ensures leak-free operation at low pressures.with specifying squeeze at the low can be calculated from end of the range is that di- b = 2.4x mensional tolerances can reduce the amount of squeeze actually Then, the peak contact stress can placed on the O-ring. For instance, be found from tolerance on the 0.070-in. thick O-ring is +/-0.003 in. In the worst case (0.067-in. thickness), this tolerance can account for 8.6% of If f ' is greater than the system the squeeze allowance, leaving only pressure, the O-ring will seal the 6.3% to be supplied by the fit joint. If f ' is less than system between parts. In other words, an pressure, the ring will leak and a undersize O-ring has less material material with a higher Young's to compress and cannot be Modulus must be specified, squeezed as tightly against the thereby increasing compressive sealing surfaces. This problem can force and contact stress.be minimized by specifying O-rings with one-half the normal Seal Friction dimensional tolerances. Such seals In low-pressure systems, seal are available from most friction can raise the required manufacturers at a premium price. actuating pressure to many times The next step in the design that available in the system.procedure is to calculate the Therefore, seal friction must be compression force developed in minimized for the system to the O-ring. This force is directly operate properly. Generally, seal related to the sealing ability of the friction force should be maintained ring and is calculated from below 20 lb to keep actuating force within reasonable limits.The friction force for an O-ring seal can be estimated from 75 1340 To use this equation, Young's Modulus must be determined first.of seal deflection to seal thickness, This value depends on material where coefficient of friction, ,x/d. Generally, the seal is designed hardness, and typical values are can change from 0.001 to over 10, to operate at the high end of the listed in Table 2. For most depending on the operating squeeze range to ensure a tight applications, a Shore A hardness of conditions.seal. But at low system pressures, 70 is sufficient; therefore, the initial When more than one O-ring is squeeze must be specified at the calculation of F is based on this used in the system, the friction low end of the range.hardness. forces from all the seals must be The squeeze values listed in From the specified squeeze and combined to determine the total Table 1 are based on nominal seal seal thickness, contact area friction force. If the calculated thickness. One problem force is greater than

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 5 of 44 20 lb, a softer seal material force must be lowered by reduc- an unlubricated seal.should be used; this lowers ing the coefficient of friction. This increase with time is Young's Modulus and compres- This factor is a complex function caused by the atomic interaction sive force. However, the change of lubricant film thickness, time, between the O-ring and its to a softer seal material must be contact stress, sliding speed, and sealing surface, which causes the made with care because a lower surface finish. two surfaces to adhere tightly.compressive force also means a Tests have shown that the The adhesive force can be quite lower contact stress. Thus, the longer a lubricated seal sits idle, high and eventually squeezes change could lower peak contact the higher its static, or most of the lubricant from under stress below system pressure, breakaway, coefficient of friction. the contact area. On start-up, the resulting in a leaky seal. Eventually, the friction adhered O-ring peels away in If a softer material lowers coefficient reaches a maximum progressive waves that break contact stress too much, friction value almost as high as that for away and reform

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 6 of 44 on the moving surface, This action shears what little lubricant is present and traps it in the rubber folds.Seal adhesion can be minimized by optimizing surface finish and lubricant viscosity. Experience has shown that the optimum surface finish is 0.4 m. This finish leaves tiny pockets that collect lubricant, making it available at startup. Too smooth a finish leaves no pockets for the lubricant, while too rough a finish causes high wear.A reciprocating seal should be lubricated with high viscosity lubricants because they produce a strong hydrodynamic film. This film resists displacement by the adhesive forces when the seal is stationary. A rotary seal, on the other hand, can be lubricated with low-viscosity lubricants because rotary motion aids development of a hydrodynamic film.Thickness of the hydrodynamic film between asperities on the O-ring and sealing surface has been calculated as 6 x 10 in. Shear of this film is the prime cause of dynamic or running friction. In general, the dynamic coefficient of friction is a function of lubricant viscosity and sliding velocity. The coefficient generally starts high, decreases to a minimum value, then increases again.Thus, running friction can be minimized by optimizing viscosity and velocity.Several tests have been run to determine the effect of material-formula modifications on seal friction. The addition of materials such as graphite, molybdenum disulphide, and PTFE sometimes reduce friction, but the reduction is more likely a result of lowering Young's Modulus than a lubricating effect. Also, the incorporation in the elastomer of high-molecular-weight waxes and oils that migrate to the surface has proved unsuccessful in

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 7 of 44 Page intentionally blank

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 8 of 44 Table 3-Glass-Transition Temperatures for lowering friction. O-ring Materials Surface treatments have been Material Transition Temperature more successful. Halogenation with(°F) chlorine or bromine reduces friction Nitrile (NBR) 34% ACN -21 by lowering the surface free energy 38% ACN -13 (and, therefore, attraction force) and Fluoro Rubber (FKM) -4 by creating lubricant pockets. Silicone (VMQ) -85 Chloroprene (CR) -40 Fluorination, although far less Ethylene Propylene (EPDM) -85 common, has similar effects.Surface treatments of PTFE-resin with material composition, binder coatings and tumble temperature, and fluid reactions. steel. Therefore, at high temperatures, treatment in molybdenum disulphide Typical values range from 0.5% to insufficient groove volume can or graphite have been used along 10% per time decade. (The time from produce expansion forces that with silicone oil dips with limited 1 to 10 min is designated as one extrude the seal into the clearances.success. Also, polymerization of decade, as is the much longer time This problem can be minimized by monomers on the O-ring surface from 1 to 10 weeks.) increasing groove dimensions to with plasma techniques offers some The result of stress relaxation is provide sufficient room for improvement; however, the that peak compressive stress expansion.techniques are costly and slow. eventually drops below system Lowering operating temperature Finally, the grafting onto the seal pressure, and the seal leaks. Thus, results in a continuous decrease in the surface of high-molecular-weight oils stress relaxation effects must be physical volume of the seal.having reactive end groups shows factored into the determination of Eventually, the seal reaches its promise for the future. material hardness and compressive so-called glass-transition temperature, stress. where it seals only along two thin Stress Relaxation Stress relaxation rates are available lines. Further reduction of from O-ring manufacturers; temperature shrinks the seal even The useful sealing life of an however, the ratings may be for a more, resulting in leakage.O-ring depends on two viscoelastic temperature or fluid condition material properties: compression set, different from that required. If the The glass-transition temperature the residual deformation of a correct data are not available, the corresponds to 100% compression material after the load is removed; stress relaxation rate can be set. At this temperature, the seal and stress relaxation, the decrease in determined from a simple can shatter like glass if subjected to stress after a given time at a constant relaxometer test, such as that a shock or impact load. Values of strain. These properties reduce the described in ASTM D-1395, or with glass transition temperature for resiliency of the seal material and a Lucas relaxometer. O-ring materials are listed in must be taken into account when Once the stress relaxation rate is Table 3. To avoid low specifying material hardness. known, the time for peak contact temperature problems, O-rings When a seal is under constant stress to equal system pressure can be should operate at temperatures compression, the initial stress decays calculated easily. In general, if the 10° to 15°F higher than those at a rate proportional to the calculated time period produces a seal listed in the table.logarithm of time. The stress life lower than 20 x 106 cycles, then a relaxation rate varies harder seal material (higher Young's Modulus and higher compressive References Nomenclature stress) must be used. 1. C.J. Derham, "Elastomeric Sealing,"Engineering, May 1977.b = Seal contact area, in.2 Temperature Effects 2. P. B. Lindley, "Engineering Design with Di = Seal inside diam, in. Natural Rubber," Malaysian Rubber Dm = Seal mean diam, in. Producers' Research Association, London, D = Seal outside diam, in. The effects of operating 1974.d = Seal thickness, in. temperature are more pronounced 3. P. B. Lindley, "Compression E = Young's Modulus, psi for low-squeeze O-rings because the Characteristics of Laterally Unrestrained F = Compressive load, lb seal has a lower tolerance for change. Rubber O-Rings," Journal of the Institution, Ff = Friction force, lb of the Rubber Industry, July/August 1967.f = Peak contact stress, psi The volumetric expansion rate for x = Seal deflection, in. rubber is about 15 times higher than

4. A. D. Roberts, "Optical Rubber," Rubber

 = Coefficient of friction that for Deuelopments, Vol. 29, No. 1, 1976.

5. A. D. Roberts, "Looking at Rubber Friction," Rubber Deuelopments, Vol. 29, No.

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 9 of 44 Table of Contents Search Next >aerospace climate control electromechanical filtration fluid & gas handling hydraulics pneumatics process control sealing & shielding Parker O-Ring Handbook ORD 5700

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Parker O-Ring Handbook Softer sealing materials, with lower hardness readings, will 2.4.4 Tensile Strength "ow more easily into the micro"ne grooves and imperfections Tensile strength is measured as the psi (pounds per square of the mating parts (the gland, bore, rod or seal "anges). This inch) or MPa (Mega Pascals) required to rupture a specimen of is particularly important in low-pressure seals because they a given elastomer material when stressed. Tensile strength is are not activated by "uid pressure. Conversely, the harder one quality assurance measurement used to insure compound Basic O-Ring Elastomers materials offer greater resistance to extrusion. Referring back uniformity. It is also useful as an indication of deterioration to the O-ring seal diagrams, Figures 1-4 through 1-7, it can of the compound after it has been in contact with a "uid for be seen that a harder O-ring will have greater resistance to long periods. If "uid contact results in only a small reduction extrusion into the narrow gap between the piston and bore. in tensile strength, seal life may still be relatively long, yet There are certain applications in which the compressive load if a large reduction of tensile strength occurs, seal life may available for assembly is limited. In these situations, Figures be relatively short. Exceptions to this rule do occur. Tensile 2-4 through 2-8 are helpful, providing compression load strength is not a proper indication of resistance to extrusion, requirements for O-rings of different hardnesses, for each nor is it ordinarily used in design calculations. However, in of the "ve standard O-ring cross-sections. dynamic applications a minimum of 1,000 psi (7 MPa) is In dynamic applications, the hardness of the O-ring is doubly normally necessary to assure good strength characteristics important because it also affects both breakout and running required for long-term sealability and wear resistance in friction. Although a harder compound will, in general, have moving systems.a lower coef"cient of friction than a softer material, the ac-tual running and breakout friction values are actually higher 2.4.5 Elongation because the compressive load required to achieve the proper Elongation is de"ned as the increase in length, expressed squeeze and force the harder material into a given O-ring numerically, as a percent of initial length. It is generally re-cavity is so much greater. ported as ultimate elongation, the increase over the original For most applications, compounds having a Shore A durom- dimension at break. This property primarily determines the eter hardness of 70 to 80 is the most suitable compromise. stretch which can be tolerated during the installation of an This is particularly true of dynamic applications where 90 O-ring. Elongation increases in importance as the diameters of durometer or harder compounds often allow a few drops of a gland become smaller. It is also a measure of the ability of a "uid to pass with each cycle, and 50 durometer compounds compound to recover from peak overload, or a force localized tend to abrade, wear, and extrude very quickly. in one small area of a seal, when considered in conjunction with tensile strength. An adverse change in the elongation Normally durometer hardness is referred to in increments of a compound after exposure to a "uid is a de"nite sign of of "ve or ten, as 60 durometer, 75 durometer, etc. not as degradation of the material. Elongation, like tensile strength, 62 durometer, 66 durometer or 73 durometer. This practice is used throughout the industry as a quality assurance measure is based on: on production batches of elastomer materials.(1) The fact that durometer is generally called out in speci"cations with a tolerance of +/-5 (i.e., 65+/-5, 70+/-5, 2.4.6 O-Ring Compression Force 90+/-5);O-ring compression force is the force required to compress an (2) The inherent minor variance from batch to batch of a O-ring the amount necessary to maintain an adequate sealing given rubber compound due to slight differences in raw line of contact. See Table 2-3 and Figures 2-4 through 2-8. It materials and processing techniques; and is very important in some applications, particularly in face-type (3) The human variance encountered in reading durometer seals where the available compression load is limited. The hardness. On a 70-durometer stock, for example, one factors that in"uence compression force for a given applica-person might read 69 and another 71. This small dif- tion, and a method of "nding its approximate magnitude are ference is to be expected and is considered to be within explained in Section III, O-Ring Applications.acceptable experimental error and the accuracy of the testing equipment.O-Ring Compression Force 2.4.3 Toughness Durometer Diameter Compression Toughness is not a measured property or parameter but rather a Range Load qualitative term frequently used to summarize the combination Less than normal Less than 25.4 mm (1") Middle third of range of resistance to physical forces other than chemical action. It Less than normal Over 25.4 mm (1") Lower half of range is used as a relative term in practice. The following six terms Over normal Less than 25.4 mm (1") Upper third of range (paragraphs 2.4.4 through 2.4.9) are major indicators of, and describe the toughness of a compound. Over normal Over 25.4 mm (1") Upper half of range Table 2-3: O-Ring Compression Force 2-10 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

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Parker O-Ring Handbook

 .070 Cross Section 90 80 40% 70 Basic O-Ring Elastomers 60 50 90 80 30% 70 60 Ha Percent Compression rd 50 ne ss A 90 Sh 80 or 20% e 70 60 50 90 80 10% 70 60 50 90 80 5%

70 60 50

 .1 .2 .3 .4 .5 .6 .7.8.91 2 3 4 5 6 7 8 9 10 2 3 4 5 6 7 8 9 100 2 3 4 5 6 7 8 9 1000 Compression Load per Linear Inch of Seal Pounds Figure 2-4: .070 Cross Section .103 Cross Section 90 80 40% 70 60 50 90 80 30% 70 60 Ha Percent Compression rd 50 ne ss A 90 Sh 80 or 20% e 70 60 50 90 80 10% 70 60 50 90 80 5%

70 60 50

 .1 .2 .3 .4 .5 .6 .7.8.91 2 3 4 5 6 7 8 9 10 2 3 4 5 6 7 8 9 100 2 3 4 5 6 7 8 9 1000 Compression Load per Linear Inch of Seal Pounds Figure 2-5: .103 Cross Section Parker Hanni"n Corporation

O-Ring Division 2-11 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 12 of 44

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Parker O-Ring Handbook

 .139 Cross Section 90 80 40% 70 Basic O-Ring Elastomers 60 50 90 80 30% 70 Ha 60 rd Percent Compression ne 50 ss A

Sh 90 ore 80 20% 70 60 50 90 80 10% 70 60 50 90 80 5%70 60 50

 .1 .2 .3 .4 .5 .6 .7.8.91 2 3 4 5 6 7 8 9 10 2 3 4 5 6 7 8 9 100 2 3 4 5 6 7 8 9 1000 Compression Load per Linear Inch of Seal Pounds Figure 2-6: .139 Cross Section .210 Cross Section 90 80 40% 70 60 50 90 80 30% 70 Ha 60 rd Percent Compression ne 50 ss A 90 Sh 80 ore 20% 70 60 50 90 80 10% 70 60 0

50 90 80 5%70 60 50

 .1 .2 .3 .4 .5 .6 .7.8.9 1 2 3 4 5 6 7 8 9 10 2 3 4 5 6 7 8 9 100 2 3 4 5 6 7 8 9 1000 Compression Load per Linear Inch of Seal Pounds Figure 2-7: .210 Cross Section 2-12 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 13 of 44

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Parker O-Ring Handbook

 .275 Cross Section 90 80 40% 70 Basic O-Ring Elastomers 60 50 90 80 30% 70 60 Percent Compression Ha 50 rd ne ss A 90 Sh 80 20% ore 70 60 50 90 80 10% 70 60 50 90 80 5%

70 60 50

 .1 .2 .3 .4 .5 .6 .7.8.91 2 3 4 5 6 7 8 9 10 2 3 4 5 6 7 8 9 100 2 3 4 5 6 7 8 9 1000 Compression Load per Linear Inch of Seal Pounds Figure 2-8: .275 Cross Section 2.4.7 Modulus 2.4.9 Abrasion Resistance Modulus, as used in rubber terminology, refers to stress at a Abrasion resistance is a general term that indicates the wear predetermined elongation, usually 100%. It is expressed in resistance of a compound. Where tear resistance essentially pounds per square inch (psi) or MPa (Mega Pascals). This is concerns cutting or otherwise rupturing the surface, abra-actually the elastic modulus of the material. sion resistance concerns scraping or rubbing of the surface.

This is of major importance for dynamic seal materials. Only The higher the modulus of a compound, the more apt it is to certain elastomers are recommended for dynamic O-ring recover from peak overload or localized force, and the bet-service where moving parts actually contact the seal material.ter its resistance to extrusion. Modulus normally increases Harder compounds, up to 90 durometer, are normally more with an increase in hardness. It is probably the best overall resistant to abrasion than softer compounds. Of course, as indicator of the toughness of a given compound, all other with all sealing compromises, abrasion resistance must be factors being equal.considered in conjunction with other physical and chemical requirements.2.4.8 Tear Resistance Tear strength is relatively low for most compounds. 2.4.10 Volume Change However, if it is extremely low (less than 100 lbs./in.)Volume change is the increase or decrease of the volume of an (17.5 kn/m) , there is increased danger of nicking or cutting the elastomer after it has been in contact with a "uid, measured O-ring during assembly, especially if it must pass over ports, in percent (%).sharp edges or burrs. Compounds with poor tear resistance will fail quickly under further "exing or stress once a crack Swell or increase in volume is almost always accompanied by is started. In dynamic seal applications, inferior tear strength a decrease in hardness. As might be surmised, excessive swell of a compound is also indicative of poor abrasion resistance will result in marked softening of the rubber. This condition which may lead to premature wear and early failure of the will lead to reduced abrasion and tear resistance, and may seal. Usually however, this property need not be considered permit extrusion of the seal under high pressure.for static applications.For static O-ring applications volume swell up to 30% can usually be tolerated. For dynamic applications, 10 or 15%swell is a reasonable maximum unless special provisions are made in the gland design itself. This is a rule-of-thumb and there will be occasional exceptions to the rule.Parker Hanni"n Corporation

O-Ring Division 2-13 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 14 of 44 Parker O-Ring Handbook is less than .1 mm (.005 inch). The three curves, repre-Compression Recovery of Three O-Ring senting three nitrile compounds, show very clearly that Compounds When Light Squeeze is Applied a good compression set resistant compound can be 100 Recovery After distinguished from a poor one only when the applied Compression of squeeze exceeds .1 mm (.005 inches).Recovery 70 Hours at 75 100°C (212°F) Most seal applications cannot tolerate a no or zero O-Ring Applications Recovery is Essentially squeeze condition. Exceptions include low-pressure Independent of air valves, for which the "oating pneumatic piston ring Sample Thickness 50 design is commonly used, and some rotary O-ring seal Percent of Original Delection applications. See the Dynamic O-Ring Sealing, Section V, and Tables A6-6 and A6-7 for more information on 25 pneumatic and rotary O-ring seal design.0 3.7 Gland Fill mm 0 0.1 0.3 0.4 0.5 The percentage of gland volume that an O-ring In. 0 0.005 0.010 0.015 0.020 Compression cross-section displaces in its con"ning gland is called gland "ll. Most O-ring seal applications call for a gland Figure 3-5: Compression Recovery of Three O-ring "ll of between 60% to 85% of the available volume with Compounds When Light Squeeze is Applied the optimum "ll being 75% (or 25% void). The reason for the 60% to 85% range is because of potential An assembled stretch greater than "ve percent is not tolerance stacking, O-ring volume swell and possible recommended because the internal stress on the O-ring thermal expansion of the seal. It is essential to allow causes more rapid aging. Over "ve percent stretch at least a 10% void in any elastomer sealing gland.may sometimes be used, however, if a shorter useful life is acceptable.3.8 O-Ring Compression Force Of the commonly used O-ring seal elastomers, the The force required to compress each linear inch of an reduction in useful life is probably greatest with nitrile O-ring seal depends principally on the shore hardness materials. Therefore, where high stretch is necessary, of the O-ring, its cross-section, and the amount of it is best to use ethylene propylene, "uorocarbon, compression desired. Even if all these factors are the polyurethane or neoprene, whichever material has the same, the compressive force per linear inch for two necessary resistance to the temperatures and "uids rings will still vary if the rings are made from different involved. compounds or if their inside diameters are different.The anticipated load for a given installation is not 3.6 Squeeze "xed, but is a range of values. The values obtained The tendency of an O-ring to attempt to return to its from a large number of tests are expressed in the bar original uncompressed shape when the cross-section charts of Figures 2-4 through 2-8 in Section II. If the is de"ected is the basic reason why O-rings make such hardness of the compound is known quite accurately, excellent seals. Obviously then, squeeze is a major the table for O-ring compression force, Table 2-3 may consideration in O-ring seal design. be used to determine which portion of the bar is most likely to apply.In dynamic applications, the maximum recommended squeeze is approximately 16%, due to friction and Increased service temperatures generally tend to wear considerations, though smaller cross-sections soften elastomeric materials (at least at "rst). Yet the may be squeezed as much as 25%. compression force decreases very little except for the hardest compounds. For instance, the compression When used as a static seal, the maximum recom- force for O-rings in compound N0674-70 decreased mended squeeze for most elastomers is 30%, though only 10% as the temperature was increased from 24°C this amount may cause assembly problems in a radial (75°F) to 126°C (258°F). In compound N0552-90 the squeeze seal design. In a face seal situation, however, compression force decrease was 22% through the a 30% squeeze is often bene"cial because recovery same temperature range.is more complete in this range, and the seal may function at a somewhat lower temperature. There is a Refer to Figure 3-6 for the following information:danger in squeezing much more than 30% since the The dotted line indicates the approximate linear extra stress induced may contribute to early seal de- change in the cross section (W) of an O-ring when terioration. Somewhat higher squeeze may be used if the gland prevents any change in the I.D. with the seal will not be exposed to high temperatures nor shrinkage, or the O.D., with swell. Hence this curve to "uids that tend to attack the elastomer and cause indicates the change in the effective squeeze on additional swell. an O-ring due to shrinkage or swell. Note that vol-umetric change may not be such a disadvantage The minimum squeeze for all seals, regardless of as it appears at "rst glance. A volumetric shrinkage cross-section should be about .2 mm (.007 inches). of six percent results in only three percent linear The reason is that with a very light squeeze almost all shrinkage when the O-ring is con"ned in a gland.elastomers quickly take 100% compression set. Figure This represents a reduction of only .003" of squeeze 3-5 illustrates this lack of recovery when the squeeze on an O-ring having a .103" cross-section (W)WARNING: These products can expose you to chemicals including carbon black (airborne and extracts), antimony trioxide, titanium dioxide, silica (crystalline),di(2-ethylhexyl)phthalate, ethylene thiourea, acrylonitrile, 1,3-butadiene, epichlorohydrin, toluenediisocyanate, tetrafluoroethylene, ethylbenzene, formaldehyde, furfuryl alcohol, glass fibers, methyl isobutyl ketone, nickel (metallic and compounds), lead and lead compounds which are known to the State of California to cause cancer; and 1,3-butadiene, epichlorohydrin, di(2-ethylhexyl)phthalate, di-isodecyl phthalate, ethylene thiourea, methyl isobutyl ketone, methanol, toluene, lead and lead compounds which are known to the State of Califormia to cause birth defects and other reproductive harm. For more information go to www.P65Warnings.ca.gov.Parker Hanni"n Corporation

O-Ring Division 3-9 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 15 of 44

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Parker O-Ring Handbook Section IV - Static O-Ring Sealing 4.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 Face Seal Glands Design Chart 4-3 . . . . . . . . . . . . . . . . . . . . . . 4-18 4.1 Surface Finishes for Static O-Ring Seals. . . . . . . . . 4-2 Static O-Ring Sealing Dovetail Grooves 4.2 Static Male and Female O-Ring Design . . . . . . . . . 4-2 Design Chart 4-4 . . . . . . . . . . . . . . . . . . . . . . 4-19 4.3 Face Type O-Ring Seals. . . . . . . . . . . . . . . . . . . . . . 4-2 Half Dovetail Grooves 4.4 Dovetail and Half-Dovetail Grooves . . . . . . . . . . . . 4-3 Design Chart 4-5 . . . . . . . . . . . . . . . . . . . . . . 4-20 4.5 Boss Seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 Static Crush Seal Grooves 4.6 Failures and Leakage . . . . . . . . . . . . . . . . . . . . . . . . 4-3 Design Chart 4-6 . . . . . . . . . . . . . . . . . . . . . . 4-21 4.7 O-Ring Glands . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 Tube Fitting Boss Seals AS5202 4.7.1 O-Ring Glands for Aerospace Design Table 4-3 . . . . . . . . . . . . . . . . . . . . . . 4-22 Hydraulic Packings and Gaskets . . . . . . . . . . . . . 4-3 Tube Fitting Boss Seals AS4395 Design Chart 4-1 A & B . . . . . . . . . . . . . . . . . 4-4 Design Table 4-4 . . . . . . . . . . . . . . . . . . . . . . 4-23 Design Table 4-1 . . . . . . . . . . . . . . . . . . . . . . . 4-5 Design Table 4-5 . . . . . . . . . . . . . . . . . . . . . . 4-24 4.7.2 O-Ring Glands for Industrial Static Seals Vacuum Seal Glands Design Chart 4-2 . . . . . . . . . . . . . . . . . . . . . . . 4-9 Design Chart 4-7 . . . . . . . . . . . . . . . . . . . . . . 4-25 Design Table 4-2 . . . . . . . . . . . . . . . . . . . . . . 4-10 Parker Hanni"n Corporation

O-Ring Division 4-1 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 16 of 44

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Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 2-001 .029 .004 .040 .109 .105 .101 .040 .044 * .103 .042 .055 002 .042 .004 .050 .142 .138 .132 .002 .053 .059 .002 * .136 .055 .070 003 .056 .004 .060 .176 .172 .162 .067 .077 * .170 .069 .083 004 .070 .005 .210 .206 .181 .081 .106 * .204 .083 005 .101 .005 .241 .237 .212 .112 .137 * .235 .114 006 .114 .005 .254 .250 .225 .125 .150 * .248 .127 007 .145 .005 .285 .281 .256 .156 .181 * .279 .158 008 .176 .005 .316 .312 .287 .187 .212 * .310 .189 009 .208 .005 .348 .343 .318 .218 .243 * .341 .220 010 .239 .005 .379 .375 .350 .250 .275 * .373 .252 011 .301 .005 .441 .437 .412 .312 .337 * .435 .314 012 .364 .005 .504 .500 .475 .375 .400 * .498 .377 013 .426 .005 .566 .562 .537 .437 .462 .560 .439 014 .489 .005 .629 .625 .600 .500 .525 .623 .502 015 .551 .007 .691 .687 .662 .562 .587 .685 .564 016 .614 .009 .754 .750 .725 .625 .650 .748 .627 017 .676 .009 .816 .812 .787 .687 .712 .810 .689 018 .739 .009 .879 .875 .850 .750 .775 .873 .752 019 .801 .009 .941 .937 .912 .812 .837 .935 .814 020 .864 .009 1.004 1.000 .975 .875 .900 .998 .877 021 .926 .009 1.066 1.062 1.037 .937 .962 1.060 .939 .093 022 .989 .010 .070 1.129 1.125 1.100 .002 1.000 1.025 .002 1.123 1.002 023 1.051 .010 +/-.003 1.191 1.187 1.162 1.062 1.087 1.185 1.064 024 1.114 .010 1.254 1.250 1.225 1.125 1.150 1.248 1.127 025 1.176 .011 1.316 1.312 1.287 1.187 1.212 1.310 1.189 026 1.239 .011 1.379 1.375 1.350 1.250 1.275 1.373 1.252 027 1.301 .011 1.441 1.437 1.412 1.312 1.337 1.435 1.314 028 1.364 .013 1.504 1.500 1.475 1.375 1.400 1.498 1.377 029 1.489 .013 1.629 1.625 1.600 1.500 1.525 1.623 1.502 030 1.614 .013 1.754 1.750 1.725 1.625 1.650 1.748 1.627 031 1.739 .015 1.879 1.875 1.850 1.750 1.775 1.873 1.752 032 1.864 .015 2.004 2.000 1.975 1.875 1.900 1.998 1.877 033 1.989 .018 2.129 2.125 2.100 2.000 2.025 2.123 2.002 034 2.114 .018 2.254 2.250 2.225 2.125 2.150 2.248 2.127 035 2.239 .018 2.379 2.375 2.350 2.250 2.275 2.373 2.252 036 2.364 .018 2.504 2.500 2.475 2.375 2.400 2.498 2.377 037 2.489 .018 2.629 2.625 2.600 2.500 2.525 2.623 2.502 038 2.614 .020 2.754 2.750 2.725 2.625 2.650 2.748 2.627 039 2.739 .020 2.879 2.875 2.850 2.750 2.775 2.873 2.752 040 2.864 .020 3.004 3.000 2.975 2.875 2.900 2.998 2.877 041 2.989 .024 3.129 3.125 3.100 3.000 3.025 3.123 3.002 042 3.239 .024 3.379 3.375 3.350 3.250 3.275 3.373 3.252 043 3.489 .024 3.629 3.625 3.600 3.500 3.525 3.623 3.502 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.4-10 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 17 of 44

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Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 044 3.739 .027 3.879 3.875 3.850 3.750 3.775 3.873 3.752 045 3.989 .027 .070 4.129 4.125 4.100 .002 4.000 4.025 .002 4.123 4.002 .093 046 4.239 .030 +/-.003 4.379 4.375 4.350 4.250 4.275 4.373 4.252 047 4.489 .030 4.629 4.625 4.600 4.500 4.525 4.623 4.502 048 4.739 .030 4.879 4.875 4.850 4.750 4.775 4.873 4.752 049 4.989 .037 5.129 5.125 5.100 5.000 5.025 5.123 5.002 050 5.239 .037 5.379 5.375 5.350 5.250 5.275 5.373 5.252 102 .049 .005 .255 .247 .224 .062 .085 * .245 .064 103 .081 .005 .287 .278 .256 .094 .116 * .276 .095 104 .112 .005 .318 .310 .287 .125 .148 * .308 .127 105 .143 .005 .349 .342 .318 .156 .180 * .340 .158 106 .174 .005 .380 .374 .349 .187 .212 * .372 .189 107 .206 .005 .412 .405 .381 .219 .243 * .403 .221 108 .237 .005 .443 .437 .412 .250 .275 * .435 .252 109 .299 .005 .505 .500 .474 .312 .338 * .498 .314 110 .362 .005 .568 .562 .537 .375 .400 * .560 .377 111 .424 .005 .630 .625 .599 .437 .463 * .623 .439 112 .487 .005 .693 .687 .662 .500 .525 * .685 .502 113 .549 .007 .755 .750 .724 .562 .588 * .748 .564 114 .612 .009 .818 .812 .787 .625 .650 .810 .627 115 .674 .009 .880 .875 .849 .687 .713 .873 .689 116 .737 .009 .943 .937 .912 .750 .775 .935 .752 117 .799 .010 1.005 1.000 .974 .812 .838 .998 .814 118 .862 .010 1.068 1.062 1.037 .875 .900 1.060 .877 119 .924 .010 .103 1.130 1.125 1.099 .002 .937 .963 .002 1.123 .939 .140 120 .987 .010 +/-.003 1.193 1.187 1.162 1.000 1.025 1.185 1.002 121 1.049 .010 1.255 1.250 1.224 1.062 1.088 1.248 1.064 122 1.112 .010 1.318 1.312 1.287 1.125 1.150 1.310 1.127 123 1.174 .012 1.380 1.375 1.349 1.187 1.213 1.373 1.189 124 1.237 .012 1.443 1.437 1.412 1.250 1.275 1.435 1.252 125 1.299 .012 1.505 1.500 1.474 1.312 1.338 1.498 1.314 126 1.362 .012 1.568 1.562 1.537 1.375 1.400 1.560 1.377 127 1.424 .012 1.630 1.625 1.599 1.437 1.463 1.623 1.439 128 1.487 .012 1.693 1.687 1.662 1.500 1.525 1.685 1.502 129 1.549 .015 1.755 1.750 1.724 1.562 1.588 1.748 1.564 130 1.612 .015 1.818 1.812 1.787 1.625 1.650 1.810 1.627 131 1.674 .015 1.880 1.875 1.849 1.687 1.713 1.873 1.689 132 1.737 .015 1.943 1.937 1.912 1.750 1.775 1.935 1.752 133 1.799 .015 2.005 2.000 1.974 1.812 1.838 1.998 1.814 134 1.862 .015 2.068 2.062 2.037 1.875 1.900 2.060 1.877 135 1.925 .017 2.131 2.125 2.099 1.937 1.963 2.123 1.939 136 1.987 .017 2.193 2.187 2.162 2.000 2.025 2.185 2.002 137 2.050 .017 2.256 2.250 2.224 2.062 2.088 2.248 2.064 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.Parker Hanni"n Corporation

O-Ring Division 4-11 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 18 of 44

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Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 138 2.112 .017 2.318 2.312 2.287 2.125 2.150 2.310 2.127 139 2.175 .017 2.381 2.375 2.349 2.187 2.213 2.373 2.189 140 2.237 .017 2.443 2.437 2.412 2.250 2.275 2.435 2.252 141 2.300 .020 2.506 2.500 2.474 2.312 2.338 2.498 2.315 142 2.362 .020 2.568 2.562 2.537 2.375 2.400 2.560 2.377 143 2.425 .020 2.631 2.625 2.599 2.437 2.463 2.623 2.439 144 2.487 .020 2.693 2.687 2.662 2.500 2.525 2.685 2.502 145 2.550 .020 2.756 2.750 2.724 2.562 2.588 2.748 2.564 146 2.612 .020 2.818 2.812 2.787 2.625 2.650 2.810 2.627 147 2.675 .022 2.881 2.875 2.849 2.687 2.713 2.873 2.689 148 2.737 .022 2.943 2.937 2.912 2.750 2.775 2.935 2.752 149 2.800 .022 3.006 3.000 2.974 2.812 2.838 2.998 2.814 150 2.862 .022 3.068 3.062 3.037 2.875 2.900 3.060 2.877 151 2.987 .024 3.193 3.187 3.162 3.000 3.025 3.185 3.002 152 3.237 .024 3.443 3.437 3.412 3.250 3.275 3.435 3.252 153 3.487 .024 3.693 3.687 3.662 3.500 3.525 3.685 3.502 154 3.737 .028 .103 3.943 3.937 3.912 .002 3.750 3.775 .002 3.935 3.752 .140 155 3.987 .028 +/-.003 4.193 4.187 4.162 4.000 4.025 4.185 4.002 156 4.237 .030 4.443 4.437 4.412 4.250 4.275 4.435 4.252 157 4.487 .030 4.693 4.687 4.662 4.500 4.525 4.685 4.502 158 4.737 .030 4.943 4.937 4.912 4.750 4.775 4.935 4.752 159 4.987 .035 5.193 5.187 5.162 5.000 5.025 5.185 5.002 160 5.237 .035 5.443 5.437 5.412 5.250 5.275 5.435 5.252 161 5.487 .035 5.693 5.687 5.662 5.500 5.525 5.685 5.502 162 5.737 .035 5.943 5.937 5.912 5.750 5.775 5.935 5.752 163 5.987 .035 6.193 6.187 6.162 6.000 6.025 6.185 6.002 164 6.237 .040 6.443 6.437 6.412 6.250 6.275 6.435 6.252 165 6.487 .040 6.693 6.687 6.662 6.500 6.525 6.685 6.502 166 6.737 .040 6.943 6.937 6.912 6.750 6.775 6.935 6.752 167 6.987 .040 7.193 7.187 7.162 7.000 7.025 7.185 7.002 168 7.237 .045 7.443 7.437 7.412 7.250 7.275 7.435 7.252 169 7.487 .045 7.693 7.687 7.662 7.500 7.525 7.685 7.502 170 7.737 .045 7.943 7.937 7.912 7.750 7.775 7.935 7.752 171 7.987 .045 8.193 8.187 8.162 8.000 8.025 8.185 8.002 172 8.237 .050 8.443 8.437 8.412 8.250 8.275 8.435 8.252 173 8.487 .050 8.693 8.687 8.662 8.500 8.525 8.685 8.502 174 8.737 .050 8.943 8.937 8.912 8.750 8.775 8.935 8.752 175 8.987 .050 9.193 9.187 9.162 9.000 9.025 9.185 9.002 176 9.237 .055 9.443 9.437 9.412 9.250 9.275 9.435 9.252 177 9.487 .055 9.693 9.687 9.662 9.500 9.525 9.685 9.502 178 9.737 .055 9.943 9.937 9.912 9.750 9.775 9.935 9.752 201 .171 .005 .139 .449 .437 .409 .187 .215 * .434 .190 202 .234 .005 +/-.004 .512 .500 .472 .002 .250 .278 .002 * .497 .253 .187 203 .296 .005 .574 .562 .534 .312 .340 * .559 .315 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.4-12 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 19 of 44

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Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 204 .359 .005 .637 .625 .597 .375 .403 .622 .378 205 .421 .005 .699 .687 .659 .437 .465 .684 .440 206 .484 .005 .762 .750 .722 .500 .528 .747 .503 207 .546 .007 .824 .812 .784 .562 .590 .809 .565 208 .609 .009 .887 .875 .847 .625 .653 .872 .628 209 .671 .009 .949 .937 .909 .687 .715 .934 .690 210 .734 .010 1.012 1.000 .972 .750 .778 .997 .753 211 .796 .010 1.074 1.062 1.034 .812 .840 1.059 .815 212 .859 .010 1.137 1.125 1.097 .875 .903 1.122 .878 213 .921 .010 1.199 1.187 1.159 .937 .965 1.184 .940 214 .984 .010 1.262 1.250 1.222 1.000 1.028 1.247 1.003 215 1.046 .010 1.324 1.312 1.284 1.062 1.090 1.309 1.065 216 1.109 .012 1.387 1.375 1.347 1.125 1.153 1.372 1.128 217 1.171 .012 1.449 1.437 1.409 1.187 1.215 1.434 1.190 218 1.234 .012 1.512 1.500 1.472 1.250 1.278 1.497 1.253 219 1.296 .012 1.574 1.562 1.534 1.312 1.340 1.559 1.315 220 1.359 .012 .139 1.637 1.625 1.597 .002 1.375 1.403 .002 1.622 1.378 .187 221 1.421 .012 +/-.004 1.700 1.687 1.659 1.437 1.465 1.684 1.440 222 1.484 .015 1.762 1.750 1.722 1.500 1.528 1.747 1.503 223 1.609 .015 1.887 1.875 1.847 1.625 1.653 1.872 1.628 224 1.734 .015 2.012 2.000 1.972 1.750 1.778 1.997 1.753 225 1.859 .015 2.137 2.125 2.097 1.875 1.903 2.122 1.878 226 1.984 .018 2.262 2.250 2.222 2.000 2.028 2.247 2.003 227 2.109 .018 2.387 2.375 2.347 2.125 2.153 2.372 2.128 228 2.234 .020 2.512 2.500 2.472 2.250 2.278 2.497 2.253 229 2.359 .020 2.637 2.625 2.597 2.375 2.403 2.622 2.378 230 2.484 .020 2.762 2.750 2.722 2.500 2.528 2.747 2.503 231 2.609 .020 2.887 2.875 2.847 2.625 2.653 2.872 2.628 232 2.734 .024 3.012 3.000 2.972 2.750 2.778 2.997 2.753 233 2.859 .024 3.137 3.125 3.097 2.875 2.903 3.122 2.878 234 2.984 .024 3.262 3.250 3.222 3.000 3.028 3.247 3.003 235 3.109 .024 3.387 3.375 3.347 3.125 3.153 3.372 3.128 236 3.234 .024 3.512 3.500 3.472 3.250 3.278 3.497 3.253 237 3.359 .024 3.637 3.625 3.597 3.375 3.403 3.622 3.378 238 3.484 .024 3.762 3.750 3.722 3.500 3.528 3.747 3.503 239 3.609 .028 3.887 3.875 3.847 3.625 3.653 3.872 3.628 240 3.734 .028 4.012 4.000 3.972 3.750 3.778 3.997 3.753 241 3.859 .028 4.137 4.125 4.097 3.875 3.903 4.122 3.878 242 3.984 .028 4.262 4.250 4.222 4.000 4.028 4.247 4.003 243 4.109 .028 4.387 4.375 4.347 4.125 4.153 4.372 4.128 244 4.234 .030 4.512 4.500 4.472 4.250 4.278 4.497 4.253 245 4.359 .030 4.637 4.625 4.597 4.375 4.403 4.622 4.378 246 4.484 .030 4.762 4.750 4.722 4.500 4.528 4.747 4.503 247 4.609 .030 4.887 4.875 4.847 4.625 4.653 4.872 4.628 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.Parker Hanni"n Corporation

O-Ring Division 4-13 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 20 of 44

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Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 248 4.734 .030 5.012 5.000 4.972 4.750 4.778 4.997 4.753 249 4.859 .035 5.137 5.125 5.097 4.875 4.903 5.122 4.878 250 4.984 .035 5.262 5.250 5.222 5.000 5.028 5.247 5.003 251 5.109 .035 5.387 5.375 5.347 5.125 5.153 5.372 5.128 252 5.234 .035 5.512 5.500 5.472 5.250 5.278 5.497 5.253 253 5.359 .035 5.637 5.625 5.597 5.375 5.403 5.622 5.378 254 5.484 .035 5.762 5.750 5.722 5.500 5.528 5.747 5.503 255 5.609 .035 5.887 5.875 5.847 5.625 5.653 5.872 5.628 256 5.734 .035 6.012 6.000 5.972 5.750 5.778 5.997 5.753 257 5.859 .035 6.137 6.125 6.097 5.875 5.903 6.122 5.878 258 5.984 .035 6.262 6.250 6.222 6.000 6.028 6.247 6.003 259 6.234 .040 6.512 6.500 6.472 6.250 6.278 6.497 6.253 260 6.484 .040 6.762 6.750 6.722 6.500 6.528 6.747 6.503 261 6.734 .040 7.012 7.000 6.972 6.750 6.778 6.997 6.753 262 6.984 .040 7.262 7.250 7.222 7.000 7.028 7.247 7.003 263 7.234 .045 7.512 7.500 7.472 7.250 7.278 7.497 7.253 264 7.484 .045 7.762 7.750 7.722 7.500 7.528 7.747 7.503 265 7.734 .045 .139 8.012 8.000 7.972 .002 7.750 7.778 .002 7.997 7.753 .187 266 7.984 .045 +/-.004 8.262 8.250 8.222 8.000 8.028 8.247 8.003 267 8.234 .050 8.512 8.500 8.472 8.250 8.278 8.497 8.253 268 8.484 .050 8.762 8.750 8.722 8.500 8.528 8.747 8.503 269 8.734 .050 9.012 9.000 8.972 8.750 8.778 8.997 8.753 270 8.984 .050 9.262 9.250 9.222 9.000 9.028 9.247 9.003 271 9.234 .055 9.512 9.500 9.472 9.250 9.278 9.497 9.253 272 9.484 .055 9.762 9.750 9.722 9.500 9.528 9.747 9.503 273 9.734 .055 10.012 10.000 9.972 9.750 9.778 9.997 9.753 274 9.984 .055 10.262 10.250 10.222 10.000 10.028 10.247 10.003 275 10.484 .055 10.762 10.750 10.722 10.500 10.528 10.747 10.503 276 10.984 .065 11.262 11.250 11.222 11.000 11.028 11.247 11.003 277 11.484 .065 11.762 11.750 11.722 11.500 11.528 11.747 11.503 278 11.984 .065 12.262 12.250 12.222 12.000 12.028 12.247 12.003 279 12.984 .065 13.262 13.250 13.222 13.000 13.028 13.247 13.003 280 13.984 .065 14.262 14.250 14.222 14.000 14.028 14.247 14.003 281 14.984 .065 15.262 15.250 15.222 15.000 15.028 15.247 15.003 282 15.955 .075 16.233 16.250 16.222 16.000 16.028 16.247 16.003 283 16.955 .080 17.233 17.250 17.222 17.000 17.028 17.247 17.003 284 17.955 .085 18.233 18.250 18.222 18.000 18.028 18.247 18.003 309 .412 .005 .832 .812 .777 .437 .472 * .809 .440 310 .475 .005 .210 .895 .875 .840 .500 .535 * .872 .503 311 .537 .007 +/-.005 .957 .937 .902 .004 .562 .597 .004 * .934 .565 .281 312 .600 .009 1.020 1.000 .965 .625 .660 .997 .628 313 .662 .009 1.082 1.062 1.027 .687 .722 1.059 .690 314 .725 .010 1.145 1.125 1.090 .750 .785 1.122 .753 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.4-14 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 21 of 44

 < Back Section Contents Table of Contents Search Next >

Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 315 .787 .010 1.207 1.187 1.152 .812 .847 1.184 .815 316 .850 .010 1.270 1.250 1.215 .875 .910 1.247 .878 317 .912 .010 1.332 1.312 1.277 .937 .972 1.309 .940 318 .975 .010 1.395 1.375 1.340 1.000 1.035 1.372 1.003 319 1.037 .010 1.457 1.437 1.402 1.062 1.097 1.434 1.065 320 1.100 .012 1.520 1.500 1.465 1.125 1.160 1.497 1.128 321 1.162 .012 1.582 1.562 1.527 1.187 1.222 1.559 1.190 322 1.225 .012 1.645 1.625 1.590 1.250 1.285 1.622 1.253 323 1.287 .012 1.707 1.687 1.652 1.312 1.347 1.684 1.315 324 1.350 .012 1.770 1.750 1.715 1.375 1.410 1.747 1.378 325 1.475 .015 1.895 1.875 1.840 1.500 1.535 1.872 1.503 326 1.600 .015 2.020 2.000 1.965 1.625 1.660 1.997 1.628 327 1.725 .015 2.145 2.125 2.090 1.750 1.785 2.122 1.753 328 1.850 .015 2.270 2.250 2.215 1.875 1.910 2.247 1.878 329 1.975 .018 2.395 2.375 2.340 2.000 2.035 2.372 2.003 330 2.100 .018 2.520 2.500 2.465 2.125 2.160 2.497 2.128 331 2.225 .018 2.645 2.625 2.590 2.250 2.285 2.622 2.253 332 2.350 .018 2.770 2.750 2.715 2.375 2.410 2.747 2.378 333 2.475 .020 2.895 2.875 2.840 2.500 2.535 2.872 2.503 334 2.600 .020 3.020 3.000 2.965 2.625 2.660 2.997 2.628 335 2.725 .020 3.145 3.125 3.090 2.750 2.785 3.122 2.753 336 2.850 .020 .210 3.270 3.250 3.215 .004 2.875 2.910 .004 3.247 2.878 .281 337 2.975 .024 +/-.005 3.395 3.375 3.340 3.000 3.035 3.372 3.003 338 3.100 .024 3.520 3.500 3.465 3.125 3.160 3.497 3.128 339 3.225 .024 3.645 3.625 3.590 3.250 3.285 3.622 3.253 340 3.350 .024 3.770 3.750 3.715 3.375 3.410 3.747 3.378 341 3.475 .024 3.895 3.875 3.840 3.500 3.535 3.872 3.502 342 3.600 .028 4.020 4.000 3.965 3.625 3.660 3.997 3.628 343 3.725 .028 4.145 4.125 4.090 3.750 3.785 4.122 3.753 344 3.850 .028 4.270 4.250 4.215 3.875 3.910 4.247 3.878 345 3.975 .028 4.395 4.375 4.340 4.000 4.035 4.372 4.003 346 4.100 .028 4.520 4.500 4.465 4.125 4.160 4.497 4.128 347 4.225 .030 4.645 4.625 4.590 4.250 4.285 4.622 4.253 348 4.350 .030 4.770 4.750 4.717 4.375 4.410 4.747 4.378 349 4.475 .030 4.895 4.875 4.840 4.500 4.535 4.872 4.503 350 4.600 .030 5.020 5.000 4.965 4.625 4.660 4.997 4.628 351 4.725 .030 5.145 5.125 5.090 4.750 4.785 5.122 4.753 352 4.850 .030 5.270 5.250 5.215 4.875 4.910 5.247 4.878 353 4.975 .037 5.395 5.375 5.340 5.000 5.035 5.372 5.003 354 5.100 .037 5.520 5.500 5.465 5.125 5.160 5.497 5.128 355 5.225 .037 5.645 5.625 5.590 5.250 5.285 5.622 5.253 356 5.350 .037 5.770 5.750 5.715 5.375 5.410 5.747 5.378 357 5.475 .037 5.895 5.875 5.840 5.500 5.535 5.872 5.503 358 5.600 .037 6.020 6.000 5.965 5.625 5.660 5.997 5.628 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.Parker Hanni"n Corporation

O-Ring Division 4-15 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 22 of 44

 < Back Section Contents Table of Contents Search Next >

Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 359 5.725 .037 6.145 6.125 6.090 5.750 5.785 6.122 5.753 360 5.850 .037 6.270 6.250 6.215 5.875 5.910 6.247 5.878 361 5.975 .037 6.395 6.375 6.340 6.000 6.035 6.372 6.003 362 6.225 .040 6.645 6.625 6.590 6.250 6.285 6.622 6.253 363 6.475 .040 6.895 6.875 6.840 6.500 6.535 6.872 6.503 364 6.725 .040 7.145 7.125 7.090 6.750 6.785 7.122 6.753 365 6.975 .040 7.395 7.375 7.340 7.000 7.035 7.372 7.003 366 7.225 .045 7.645 7.625 7.590 7.250 7.285 7.622 7.253 367 7.475 .045 7.895 7.875 7.840 7.500 7.535 7.872 7.503 368 7.725 .045 8.145 8.125 8.090 7.750 7.785 8.122 7.753 369 7.975 .045 8.395 8.375 8.340 8.000 8.035 8.372 8.003 370 8.225 .050 8.645 8.625 8.590 8.250 8.285 8.622 8.253 371 8.475 .050 8.895 8.875 8.840 8.500 8.535 8.872 8.503 372 8.725 .050 9.145 9.125 9.090 8.750 8.785 9.122 8.753 373 8.975 .050 9.395 9.375 9.340 9.000 9.035 9.372 9.003 374 9.225 .055 9.645 9.625 9.590 9.250 9.285 9.622 9.253 375 9.475 .055 9.895 9.875 9.840 9.500 9.535 9.872 9.503 376 9.725 .055 10.145 10.125 10.090 9.750 9.785 10.122 9.753 377 9.975 .055 .210 10.395 10.375 10.340 .004 10.000 10.035 .004 10.372 10.003 .281 378 10.475 .060 +/-.005 10.895 10.875 10.840 10.500 10.535 10.872 10.503 379 10.975 .060 11.395 11.375 11.340 11.000 11.035 11.372 11.003 380 11.475 .065 11.895 11.875 11.840 11.500 11.535 11.872 11.503 381 11.975 .065 12.395 12.375 12.340 12.000 12.035 12.372 12.003 382 12.975 .065 13.395 13.375 13.340 13.000 13.035 13.372 13.003 383 13.975 .070 14.395 14.375 14.340 14.000 14.035 14.372 14.003 384 14.975 .070 15.395 15.375 15.340 15.000 15.035 15.372 15.003 385 15.955 .075 16.375 16.375 16.340 16.000 16.035 16.372 16.003 386 16.955 .080 17.375 17.375 17.340 17.000 17.035 17.372 17.003 387 17.955 .085 18.375 18.375 18.340 18.000 18.035 18.372 18.003 388 18.955 .090 19.373 19.375 19.340 19.000 19.035 19.372 19.003 389 19.955 .095 20.373 20.375 20.340 20.000 20.035 20.372 20.003 390 20.955 .095 21.373 21.375 21.340 21.000 21.035 21.372 21.003 391 21.955 .100 22.373 22.375 22.340 22.000 22.035 22.372 22.003 392 22.940 .105 23.360 23.375 23.340 23.000 23.035 23.372 23.003 393 23.940 .110 24.360 24.375 24.340 24.000 24.035 24.372 24.003 394 24.940 .115 25.360 25.375 25.340 25.000 25.035 25.372 25.003 395 25.940 .120 26.360 26.375 26.340 26.000 26.035 26.372 26.003 425 4.475 .033 5.025 5.000 4.952 4.500 4.548 4.996 4.504 426 4.600 .033 5.150 5.125 5.077 4.625 4.673 5.121 4.629 427 4.725 .033 .275 5.275 5.250 5.202 .004 4.750 4.798 .004 5.246 4.754 .375 428 4.850 .033 +/-.006 5.400 5.375 5.327 4.875 4.923 5.371 4.879 429 4.975 .037 5.525 5.500 5.452 5.000 5.048 5.496 5.004 430 5.100 .037 5.650 5.625 5.577 5.125 5.173 5.621 5.129 431 5.225 .037 5.775 5.750 5.702 5.250 5.298 5.746 5.254 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.4-16 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 23 of 44

 < Back Section Contents Table of Contents Search Next >

Parker O-Ring Handbook Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max. (Continued)A A-1 B B-1 C D G Groove Dia. Tube OD Throat Dia.Groove Width Bore Dia. Groove Dia. Plug Dia.Static O-Ring Sealing O-Ring (Male Gland) (Female Gland) (Female Gland) (Male Gland) (Male Gland) (Female Gland)Size Dimensions Parker Mean +.002 +.000 +.000 +.001 +.005 No. 2- ID +/- W OD (Ref) -.000 -.000 + -.002 +.000 - .001 -.000 -.000 432 5.350 .037 5.900 5.875 5.827 5.375 5.423 5.871 5.379 433 5.475 .037 6.025 6.000 5.952 5.500 5.548 5.996 5.504 434 5.600 .037 6.150 6.125 6.077 5.625 5.673 6.121 5.629 435 5.725 .037 6.275 6.250 6.202 5.750 5.798 6.246 5.754 436 5.850 .037 6.400 6.375 6.327 5.875 5.923 6.371 5.879 437 5.975 .037 6.525 6.500 6.452 6.000 6.048 6.496 6.004 438 6.225 .040 6.775 6.750 6.702 6.250 6.298 6.746 6.254 439 6.475 .040 7.025 7.000 6.952 6.500 6.548 6.996 6.504 440 6.725 .040 7.275 7.250 7.202 6.750 6.798 7.246 6.754 441 6.975 .040 7.525 7.500 7.452 7.000 7.048 7.496 7.004 442 7.225 .045 7.775 7.750 7.702 7.250 7.298 7.746 7.254 443 7.475 .045 8.025 8.000 7.952 7.500 7.548 7.996 7.504 444 7.725 .045 8.275 8.250 8.202 7.750 7.798 8.246 7.754 445 7.975 .045 8.525 8.500 8.452 8.000 8.048 8.496 8.004 446 8.475 .055 9.025 9.000 8.952 8.500 8.548 8.996 8.504 447 8.975 .055 9.525 9.500 9.452 9.000 9.048 9.496 9.004 448 9.475 .055 10.025 10.000 9.952 9.500 9.548 9.996 9.504 449 9.975 .055 10.525 10.500 10.452 10.000 10.048 10.496 10.000 450 10.475 .060 11.025 11.000 10.952 10.500 10.548 10.996 10.504 451 10.975 .060 11.525 11.500 11.452 11.000 11.048 11.496 11.004 452 11.475 .060 12.025 12.000 11.952 11.500 11.548 11.996 11.504 453 11.975 .060 12.525 12.500 12.452 12.000 12.048 12.496 12.004 454 12.475 .060 .275 13.025 13.000 12.952 .004 12.500 12.548 .004 12.996 12.504 .375 455 12.975 .060 +/-.006 13.525 13.500 13.452 13.000 13.048 13.496 13.004 456 13.475 .070 14.025 14.000 13.952 13.500 13.548 13.996 13.504 457 13.975 .070 14.525 14.500 14.452 14.000 14.048 14.496 14.004 458 14.475 .070 15.025 15.000 14.952 14.500 14.548 14.996 14.504 459 14.975 .070 15.525 15.500 15.452 15.000 15.048 15.496 15.004 460 15.475 .070 16.025 16.000 15.952 15.500 15.548 15.996 15.504 461 15.955 .075 16.505 16.500 16.452 16.000 16.048 16.496 16.004 462 16.455 .075 17.005 17.000 16.952 16.500 16.548 16.996 16.504 463 16.955 .080 17.505 17.500 17.452 17.000 17.048 17.496 17.004 464 17.455 .085 18.005 18.000 17.952 17.500 17.548 17.996 17.504 465 17.955 .085 18.505 18.500 18.452 18.000 18.048 18.496 18.004 466 18.455 .085 19.005 19.000 18.952 18.500 18.548 18.996 18.504 467 18.955 .090 19.505 19.500 19.452 19.000 19.048 19.496 19.004 468 19.455 .090 20.005 20.000 19.952 19.500 19.548 19.996 19.504 469 19.955 .095 20.505 20.500 20.452 20.000 20.048 20.496 20.004 470 20.955 .095 21.505 21.500 21.452 21.000 21.048 21.496 21.004 471 21.955 .100 22.505 22.500 22.452 22.000 22.048 22.496 22.004 472 22.940 .105 23.490 23.500 23.452 23.000 23.048 23.496 23.004 473 23.940 .110 24.490 24.500 24.452 24.000 24.048 24.496 24.004 474 24.940 .115 25.490 25.500 25.452 25.000 25.048 25.496 25.004 475 25.940 .120 26.490 26.500 26.452 26.000 26.048 26.496 26.004 This groove width does not permit the use of Parbak rings. For pressures above 103.5 Bar (1500 psi), consult Design Chart 4-2 for groove widths where back-up rings must be used.

These designs require considerable installation stretch. If assembly breakage is incurred, use a compound having higher elongation or use a two-piece piston.

Design Table 4-2: Gland Dimensions for Industrial O-Ring Static Seals, 103.5 Bar (1500 psi) Max.Parker Hanni"n Corporation

O-Ring Division 4-17 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 24 of 44

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Parker O-Ring Handbook Face Seal Glands For Internal Pressure (outward pressure direction) dimension the groove by its outside diameter (HO ) and width:Static O-Ring Sealing (HO ) = Mean O.D. of O-ring L (see Table 4-1)Tolerance = Minus 1% of Mean O.D., but not more than

 -.060 For External Pressure (inward pressure direction) dimension the groove by its inside diameter (Hi) and width:

(H)i = Mean I.D. of O-ring (see Table 4-1)Tolerance = Plus 1% of Mean I.D., but not more than

 +.060 0° to 5°*

(Typ.) Break Cor ners Section W-W Approx. .005 RA D. W W X .005 Max.R Groove W Surface finish X:63 63 L Depth(= Gland Depth)I.D.32 for liquids X 16 for vacuum and gases .003 Max.G Gland Detail Finishes are RMS values (Refer to Design Chart 4-3 below)O-Ring Face Seal Glands These dimensions are intended primarily for face type O-ring seals and low temperature applications.O-Ring G Size W L Groove Width R Parker Cross Section Gland Squeeze Vacuum Groove No. 2 Nominal Actual Depth Actual % Liquids and Gases Radius 004 .050 .013 19 .101 .084 .005

 .070 +/-.003 through 1/16 to to to to to to (1.78 mm) 050 .054 .023 32 .107 .089 .015 102 .074 .020 20 .136 .120 .005 .103 +/-.003 through 3/32 to to to to to to (2.62 mm) 178 .080 .032 30 .142 .125 .015 201 .101 .028 20 .177 .158 .010 .139 +/-.004 through 1/8 to to to to to to (3.53 mm 284 .107 .042 30 .187 .164 .025 309 .152 .043 21 .270 .239 .020 .210 +/-.005 through g 3/16 to to to to to to (5.33 mm) 395 .162 .063 30 .290 .244 .035 425 .201 .058 21 .342 .309 .020 .275 +/-.006 through 1/4 to to to to to to (6.99 mm) 475 .211 .080 29 .362 .314 .035 .276 .082 22 .475 .419 .030 Special .375 +/-.007 3/8 to to to to to to (9.52 mm) .286 .106 28 .485 .424 .045 .370 .112 22 .638 .560 .030 .500 +/-.008 Special 1/2 to to to to to to (12.7 mm) .380 .138 27 .645 .565 .045 Design Chart 4-3: Design Chart for O-Ring Face Seal Glands 4-18 Parker Hanni"n Corporation

O-Ring Division 2360 Palumbo Drive, Lexington, KY 40509 Phone: (859) 269-2351
Fax: (859) 335-5128 www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 25 of 44 96 VALVE TRIM Application tact the seat joint as the seating load is

1) Simpler to apply, because of fewer applied by the actuator.

restrictions, such as flow direction, If the plu~ strike~ the seat joint slightly air supply, orifice size, pressure drop, cocl~ed, it will remain co*cked until a higher etc. seating load causes it to jump sideways

2) Fewer requirements for positioners. as it slides down into the taper and slams

3) A hydraulic snubber is never re- into full joint contact. This deforms the quired. seat, causing leakage. In time, the plug

4) Quick change trim. indenture will extend to form a new off center, nearly 360

seat contact. Above SEATING, SEALING AND LEAKAGE a certain plug co*cking angle, the plug The three problems discussed in this will _not jump into place regardless of loadmg; therefore, pre-guiding and align-secUon include: ment of the plug before seating, is neces-Seating - The alignment and contact of the plug with the seat, in- sary.

cluding joint design and load- The two principal types of mis-alignment ing. are Concentric as shown in Figure 68( a)Sealing - The parameters of metal fin- and Axial as shown in Figure 68(b ).ish, joint width, and metal yielding which lead to tight Alignment sealing. Alignment of the plug on the seat for a Leakage - The amount of leakage that single ported, top and bottom guided may be expected for different ".alve mvolves concentric alignment of sealing parameters and joint eight components and axial alignment designs. of eleven, fit combinations; plus consider-Tight sealing is becoming of greater ation of the operating clearances.importance to control valve users, now One can readily see the precision ma-that improvements in designs of both chining, required of the control valve valve trim and actuators allow tight shut- manufacturer to maintain alignment of off. One valve may be used for both stop these parts. Each part must have an a1;1d throttling service at a cost saving. assembly or sliding clearance which allows Diaphragm control, valve actuators with a .minute horizontal axial shift and a very 15 to 35 psi air supplies do not develop slight co*ck. The flexibility of the stem is the high seating forces used in stop valve sufficient to allow the plug to move into designs with manual or automatic opera- true seat joint contact with light, initial tion. If they did, they would be bulky seat-contact loading.and would be sluggish in their speed of stem movement for throttling action. A small amount of leakage is to be ex- Seat Joint Design pected, because of the lower seating forces.Manufacturers normally rate their valves Flat joints, normal to the plug axis, for maximum leakage as follows: are used on some low pressure stop valves and soft seated control valves. It is not Double Seated Valves - practical to manufacture them for high

 <0.1% Cv maximum! pressure service. As discussed later, tighter sealing occurs through a sliding and bur-Single Seated Valves - nishing action of a tapered joint. Tapered <0.01% Cv maximuml joints turn the fluid gradually and are the best _for high pressure drop and for erosive Development of the springless piston service. The control valve industry uses actuator, air loaded on both sides and joints from 15° to 45°. Smaller angles using a much higher supply pressure would begin fo form sticking tapers and (100-150 psi), coupled with a positioner, larger angles give too much of a poppet led to higher seating forces and tighter effect, when cracked open at high pres-shut-off. Single seated valves can now be sure differentials, which would cause un-sealed, drop-tight to high pressure drops. stab~angeiJrf"'Q@D.g. Itfsummary:

Some cage balanced valve designs also ---ry-45

seiifarigleslind their best appli-shut off drop-tight with small diaphragm cations in either normally open or actuators. normally closed valves.

2) 30~ seat angles are a compromise Seating between high seating forces and To make a seal, a plug must first be streamlined flow, at law lifts, for low perfectly aligned and then must fully con- erosion service.

1Seat leakage flow lslaminarratherthanturbul~nt and the Ov formula is not applicable; therefore, this is simply a means of specifying an amount of acceptable leakage.

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 26 of 44 SEATING, SEALING AND LEAKAGE 97 4.SEAT PORT 4_PLUG AT SEAT CONTACT co*ckING ANGLE SEAT DIAMETER AT PLUG CONTACT (a)(b)NON CONCENTRIC co*ckED AX IS Figure 68. PLUG MISALIGNMENT WITH SEAT.

3) 15 ° seat angles are best applied to (0.01% Cv maximum); willseal3000 high pressure drop, erosive service. QSi :gresstire drop on 0.015 in. wiam,

Jo° Joint of316SS1.

Seat Joint Width 4) 300 lb.fin. - Very high pressure drop Seat joint width is a balance of adding service; drop-tight (will seal 6000

 ~ on O. 025 in. width, 20° joint of length to the flow path to increase flow 440-C SS, hardness 55 Re),

resistance vs. reduction of the seating force for a given actuator seating load. A cer- 5) 600 lb.fin. - Extremely high pres-tain minimum width is essential to es- sure service.tablish a tight seal; however, the joint must have sufficient backup strength to The apparent compressiveseatingstress-support the compressive seating load and es on joints described in items 3) and 4) must be wide enough to prevent inden- above, are 13,000 psi and 35,000 psi tion of the plug. Narrow joints are much respectively, which is well below the yield tighter than wide jomts, provided they point of the given trim materials. Thes~exceedthe minimum wid th requtrements. are the stresses normal to the joint. Elasto<--

 ~ and plastic yielding is occuring at the ., 1iignpoints of each surface making a tor-Seat Joint loading Seat joint loading is usually expressed as pounds of force per linear inch of tuous flow path for leakage restriction. er, By contrast, stop-valve seat loading with ( !ht-W mean seat joint circumference. Loading ~ard-faced seats in steam service may be 11,._

may vary from 25-600 lb.finch as given 0,000 psi, apparent stress, or four times ~below: a nominal contr<;>! valve, seating load of

1) 25 lb.fin. - Low pressure drop ser- 100 lb/linear inch for a diaphragm ac-vice; leak-tight shut-offis not required; tuator.

metal-to-metal joint. To obtain the best circular seat-joint

2) 50 lb.fin. - IModerate pressure drop contact at low stem loading use:

service; slight leakage expected (0.1%Cv maximum). 1) Cage guided trim with the seat in-

3) 100 lb.fin. - High pressure drop tegral with the cage. Horizontal and service; nearly 'drop-tight service axial alignment for seating involves

(1 100 lbs/linear Inch vertical seating force Is equivalent to a force of 200 lbs/linear inch normal to a 30

joint ( 100 lb./sln 30" ). The compressive seating stress Is 200 lb/0.015 ln.x 1 in.= 13,300 psL

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 27 of 44 98 VALVE TRIM only two parts in sliding contact. Trim Sealing Alignment of regular trim must be transferred from the seat to the body, Tight sealing requires the yielding of to the bonnet, to the guide, to the one material into the surface "waviness" plug, and back to the seat. and surface roughness of the other, as

2) Cage guided trim with the seat aligned illustratl!d by Figure 70, to block direct by the cage. leakage paths; thus, making these paths

3) A seat that is integral with body; or long and tortuous. Compressive stresses

4) Extend a thin flexible seat lip above are far below those that would yield the the means of seat retention as shown entire joint; therefore, the contact area is in Figure 29. apparent. Actually, only the peaks ofeach A radial expandable, seat ring design is surface are in contact and the concentra-illustrated by Figure 69. In this design the tion of force may then exceed the yield seating angle creates a large radial com- and will plastically deform the high spots ponent of force which expands the seat on each valve closure. Additional closures ring against the retaining collar. The require a higher seating load to achieve spring-out action of the non-circular ring the same degree of tightness, until the allows a near-perfect joint contact giving wear particles are formed and conditions drop-tight shut-off in severe thermal cycle tend to stabilize.

service. This design has been successfully used on pressure drops to 4,400 psi. The Tapered joints provide for a sliding and large flow, entrant passage also makes burnishing action as contact is made and the valve suitable for erosive service. loading occurs. This gives a tighter initial seal than a perpendicular contact and the seal remains tighter with repeated closures.The minimum width of a joint to seal gas to 1 x 10-7 cc/s/linear inch, maxi-mum leakage, is 0.04 inch. Wider joints~with the same surface finish and loading will not seal tighter. This width insures sufficient high point contact to form an adequate flow resistance path as shown by the graph of Figure 71.Extra " super-finishing" of seat joints is unnecessary for tight sealing, because as the joint opens and closes, wear parti-cles ball up on the surface, quickly re-turning it to a rougher finish. Also, some fluid contaminants tend to remain in the joint on closure and indent the surfaces.Figure 69. RADIALLY EXPANDABLE SEAT Excessive lapping generally either reduces RING DESIGN. seat tightness by increasing the actual contact area, thus, reducing the unit com-Court9SY Conoflow Corporation pressive seating stress provided by afixed actuator seating force or it destroys the original surface geometry.f - - - -- - su.u COMl'1tsno IHGHHEU HUG~n( Ht MICll:OIKCH) ,,,-t-+c-~c----sc--1.----,-~u u or IIOIIGH"(S$CO"((KJII I( UOUT PUIG ANO t hT UIS )( UY Of *ovCffM ss CONctNUIC auut PUC AMO SEU UIS>Wl'f1Nl$S HEICHI IM *,t MICIOIHCHJ fa) Under Light Seating Load lb) Under Heavy Seating Load Figure 70. MATING OF SEAT JOINT SURFACES.

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 28 of 44 SEATING, SEALING AND LEAKAGE 99 I

 \; W: .04 I --

 .03 .06 .09 .12 IALLED PARTICLES Of WIDTH OF JOINT ROUGHNESS PEUS SOFTER SEATING MATERIAL ARE REOUCED WELOEO TO HARDER SURFACE OR FREE IN JOINT. ALSO FLUIO Figure 71. MINIMUM JOINT WIDTH CONTAMINANT PARTICLES ARE IN THE JOINT.

FOR A TIGHT SEAL.(Leakage rates for 14.7 psi 6. P helium on a flat Figure 72. JOINT SURFACE AFTER circular joint.) REPEATED CLOSURES.Consolidated graph teken from Refflrence No. 48.1000,-------~-------------SUPER FINISH

WILL NOT HOLD UP

C,;I 010 90 100 1000 LOC WAVINESS HEIGHT IICROINCHES

NOHAL TO LEAKAGE PATH Figure 73. DEGREE OF SEAT JOINT FINISH VS. METHOD OBTAINING FINISH.

Chart presented in "Investigation of Leakage and Sealing Parameters". Paul Bauer, ITT Research Ji,stitute, Technical Report AFRPL*TF 153, August, 1965.L

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 29 of 44 100 IALSI EHGINEERING, INC. VALVE TRIM The following comparison illustrates the and erosive fluids at 1000 to 4000 psi relative seating force to obtain same de- pressure drop. Few stop valves could gree of tightness vs. the seat joint finish- match this performance and remain tight.ing method ( concentric lay of finishing It is customary to expect a slight degree pattern). Refer to Figures 72 and 73. of leakage and, in general control valve service applications, this has presented Finishing Method Relative Seating Force few problems. Specified permissible leak-age should be based on process consider-diamond burnished 1.0 ations (hazards resulting from leakage( after grinding) after emergency shut off, etc.). Tight shut off should be specified only when neces-lathe turned 1. 3 sary, because a higher quality more ex-pensive _valve is required.ground 2.5 Liquid Leakage is to a great extent 2.9 affected by surface tension and the parti-lapped (excessive) cular fluid wetting the joint surface, be-(90% of apparent area of)(contact actually mated) fore the new fluid attempts to enter. For precise leak testing, the joint and body should be free of all traces of oil, which Trim Leakage would preferentially wet the metal joint surfaces. Water leakage rates of an oil-Definitions Relative to Trim Leakage: free joint are higher. Oil clinging to the roughness, blocks some leakage paths, Drop tight, to be meaningful, must be causing a higher capillary displacement specified in terms of the maximum allowed pressure to establish flow of the test fluid.number of drops per unit of time ( drops/minute, cc/hour or no visible drop).Bubble tight, to gas, should be speci-fied as the maximum allowed number of Seat Leakage Specifications bubbles of a given size per minute (usually 1/8 in. diameter). Single Seated Globe Valves Zero leakage is defined as 1 x 10-8 A maximim allowable leakage of 0.01% of cc/s or about 0.3 cc/year (helium leak rate Rated Cv is often based on the nominal or at standard conditions). Zero leakage is catalog listed actuator size for the given valve often specified in critical service and re- tested with 50psi of air across the seat joint.quires very careful joint design, material selection, finish control, and sufficient seat-ing force on narrow joints. It is practical Double Seated Globe Valves only for small valves, at extra cost, and may last for only a few closing cycles. These valves are specified to have a max-Stop ualue maximum leakage rates are imum leakage rate of 0.5% to 0.1% ofrated Cv given in the Valve Manufacturers Stan- depending upon quality purchased. They are dardization Society, SP-61 as: tested with 50 psi of air across the seat joint.

1) Water tests at AP= Body CWP rat- See ISA standard 39.1.

ing (10 cc/hour/inch of valve pipe size or about 3/drops/minute). Extra Tight Shut Off for

2) Air tests at AP=80 psi (0.1 SCFM/ Single Seated Globe Valves inch of valve pipe size). A water test is often conducted at the differ-Leakage Specifications, to be meaningful and allow comparison, should include test ential pressw-e rating assigned to the valve fluid, temperature, pressure, pressure by the manufacturer and. which is based upon drop, seating force and duration of test. actuator size, air loading, spring force and direction of leakage across the plug (either Control Valve Leakage over or under the plug). Maximum leakage may be specified as 0.0005 cubic centimeters Properly designed control valves can of water per minute per inch of valve seat achieve stop valve tightness and maintain orifice diameter (not pipe size of valve end) it throughout a long service life before per psi pressw-e drop. Example: A valve hav-trim replacement; particularly with cage ing a 4 inch seat orifice and tested to 2000 psi guided, balanced trim having elastomer plug-to-cage seals. The control valve, how- differential pressure would have a maximum ever, is expected to throttle and often shuts water leakage rate of 4cdminute. Leakage off much more frequently than stop valves. may be checked with a gas instead of a liquid.

For example, some dump valves may have The maximum allowable rate is often from 4000 to 7000 opening and closing specified at 6 x 10-* cubic centimeters per sec-cycles per day, hanaling high pressure ond per inch of seat diameter.

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 30 of 44 SEATING, SEALING AND LEAKAGE 101 Soft Seats in Globe Valves These may be leak tested in accordance with SAMA Standard#PMC23.2c. Under this h standard valve actuators are sized to provide a minimum stem force of 150 pounds per inch of seat diameter over and above the plug un-balance force created by the maximum rated differential pressure. The test pressure using air is the maximum rated differential pres- for gases the leakage formula becomes sure or 300psi, whichever is the least, but not to exceed the maximum operating pressure at 1r h 3 2 2](ro+ ri) (P1 - P2 )ambient temperature.Elastomer Lin*e d Butterfly Valves Qo= [ 12µ (r0 - q) Po

X These are often tested with 50psi of air as a minimum, or maximum differential operat-ing pressure across the disc. With the down-stream side of the disc covered with water, the maximum allowable leakage rate may be specified at one bubble of air in ten seconds, AO= molecular mean free path at stan-per inch of vane diameter. dard conditions.

E = correction factor, 0. 9 for a single gas and 0. 66 for a mixture.Elastomer Sealed Ball and Plug Valves They are usually bubble tight to their rated differential pressure. Metal seated valves The problem is that the leakage test have relatively high leak rates compared to gives no indication of whether the leak globe style valves. One exception is a rotary path is one large scratch or the sum of leakage through millions of tiny tortor-cam type plug valve with leakage rates com- ous paths. Refinishing the joint may eli-parable with globe valves. minate the first cause, but the condition may already be at the practical limit of Theoretical Leakage Formula seal tightness for the latter.The equation relating the fluid proper- Type Of Gas Flow Through Seat Joint ties, flow path geometry, and flow rate is: Leakage Pattern The following is a summary of charac-2 2 -3 teristics for various leak conditions.

 'Ir (p -P )h 1 2 Minimum Restricting Dimensio n in Plug Position Type of Flow Leakage Path where: cracked open nozzle flow >0. 005 in.

h = uniform channel clearance *2 seating load turbulent 0. 0005 to Po = pressure at standard conditions build up zone channel flow 0.005 in.P2 = exit fluid pressure inlet fluid pressure valve seated and laminar flow 0.0001 to Pt = leakage begins 0.0005 In.Qo = volume rate of flow at standard conditions joint surface transition flow 0.000001 to ro, q = outside and inside radii of waviness, then (molecular and 0.0001 in.joint sealing area roughness provides laminar)µ = absolute viscosity leak paths as deformation begins When the terms are rearranged, the elastic and plastic molecular flow <0.000001 in.uniform channel clearance is: yielding of joint has closed large paths 1 Taken from Reference 50, the formula applies to a flat seat joint.2 This i~ the theoretical separation of two truly plane surfaces to give an equivalent rate of leakage caused by channels, imperfections, etc.

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 31 of 44 7102 VALVE TJUM (di Lightly restrained O-rinos for low pressure drop, installed on double V-port plug.ia} Nylon seal retained in plug.Courtesy Fisher Controls Company ib) TFE seal in pluo of cage guided, balanced valve.Courtesy Fisher Controls Company (e) O-rlng in dove-tail groove for higher pressure drop.Vent holes relieve pressure under 0-ring retaining it in the groove as the seat join t cracks open. Holes are small to prevent seal extrusion.ELASTOMER IEAL RING PRIMlRY SEAL (f) Soft seat design used in boiler feedwate, pump recirculation anticavitation low noise valve.Handles fluids to 475° F and 6000 psig Inlet pressure.ic) Elastomer seal in seat ring of split body valve. Courtesy of M asoneilan lnternationsl, Inc.Figure 74. SOFT SEAT RETENTION .

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 32 of 44 SEAT ATTACHMENT 103 When sealing to the point of achieving A secondary advantage of soft sealing molecular flow, the lighter, smaller and is that the seal, once compressed, is free higher velocity molecules, such as helium to re-expand and to follow seat distortion and hydrogen, will have the highest leak- as the pressure drop loading increases.age rates. To do this, the sealing material should have a rapid recovery rate upon removal of the load. This action will occur only at Determination of Seat Leakage Rate resilient temperatures.Determining the Seat Leakage Rate is not Hard materials such as nylon, in thin as simple as it may appear. Fluids are sections carefully restrained, can handle subject to thermal expansion and contrac- pressure drops of several thousand psi; tion during tests and air may go in or whereas, TFE materials are readily out of solution, causing volume changes abraded.in the downstream measuring system. Soft seals are useful for sealing where F or accurate testing, maintain the valves contaminants or solid material are trapped and the fluids at an equal temperature in the closed joint. Material as hard as and air dissolved in water at equilibrium 60D in a raised-seal beadform is capable conditions. of sealing 0.01 in. particles, bubble-tight, Helium leak detection requires a clean to 1500 psi pressure drop.background with large amounts of fresh The softer the seal, the better its abrasive air for maximum sensitivity. resistance, up to the point where it is The following is a summary of the damaged by pressure drop forces.sensitivity of various test methods, given Large volume, high pressure blowdown in cc/s at atmospheric conditions. systems are necessary to adequately test elastomer seals and retention means for air bubbles in water 1 x 1 o-3 to 1 x 10-4 cc/ s pressure drop strength. Elastomers, under( also air and high loading, tend to act as a fluid and soap bubbles) extrusion may occur unless the load is limited or unless a metal-to-metal stop is thermal detectors 1 x 10-4 cc/s used. Some joint designs allow soft sealing halogen detectors 1 x 1 o-5 cc/s first, followed by a metal-to-metal closure as a secondary seal in case of soft seal mass spectrometer rupture.using "sniffer" l x 10-6 to 1 x 10-8 cc/s The material properties to beconsidered (helium leak pick up probe) in the selection of a soft seat are:

1) Fluid compatibility including, swell-ing, loss of hardness, permeability, degradation;

2) Hardness; Soft Seating 3) Permanent set;

4) Rate of recovery upon removal of Resilient composition seali_n g materials load; are used to obtain bubble tight sealing with 5) Tensile and compressive strength; a small actuator force. Compressive seat- 6) Distortion before rupture; ing stresses are such, that the material is elastically deformed into the surface 7) Modulus of elasticity.

roughness of the mating metal part, to Rarely do the physical properties given block all leak paths. The permeability for sealing materials relate to the actual of the material, to the fluid, is the source conditions of loading and strain in valve of a very minor leakage. seals. There is no substitute for thermal Materials which are too soft, or that tend and blow-down testing as the means to to cold-flow (creep) under load, may be prove material, seal configuration and stiffened with fillers such as glass. When joint design.used in thin sections, and adequately re-tained, the cold-flow or permanent-set SEAT ATTA~HMENT problems may be eliminated. AND SEALING TO BODY Seals must be carefully restrained against rupture and blow-out by differen- Seat attachment and sealing to the body tial pressure. Several designs of soft-seat is a major consideration of valve sealing, retention are illustrated by Figure 74. equal in importance to joint seal, bonnet The bonding of seats to metal parts is an seal, and stem seal. Lack of seat-to-body aid, but not a total solution, because sealing gives a continuous leak, often bonds are subject to thermal shock crack- blamed on the seat-to-plug joint. In high ing and to degradation. Sufficient pressure pressure and/or steam service, leakage drop will rupture the bonding material. behind the seal will actually erode through

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Appendix A Document 3960C, Rev. 0, Attachment 3 Page 38 of 44 O-ring Seal Design Best Practices 12-15-12 Rev. 1 1 of 7

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 39 of 44 Table of Contents 1.0 O-RING SEALS - THEORY AND DESIGN PRACTICES ......................................................................... 3 2.0 ANALYSIS OF O-RING SEAL DESIGNS ................................................................................................... 6 3.0 APPENDIX ................................................................................................................................................... 7 2 of 7

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 40 of 44 1.0 O-RING SEALS - THEORY AND DESIGN PRACTICES Theory:An o-ring seal consists of an o-ring and a properly designed gland which applies a predictable deformation to the o-ring. The gland is basically a groove dimensioned to a certain height H and width W (Figure 1) to allow a fixed compression of the o-ring when the gland flanges make metal to metal contact. It is also oversized volumetrically such to allow accommodation of the o-ring as it flows under compression. Unlike gaskets which seal just by the resiliency of the material under mechanical compression of the joint, an o-ring can provide a seal both through the resiliency of the pre-compressed material and the pressure activation of the seal. The pre-compression of the o-ring applies a calculated mechanical contact stress or pressure at the o-ring contacting surfaces in the gland. As the o-ring seal is pressurized or activated the pressure on the o-ring further increases the contact stress on the o-ring contacting surfaces of the gland as the o-ring moves or flows toward the low pressure side. This means the pressure of the contained fluid transfers through the essentially incompressible o-ring material, and the contact stress rises with increasing pressure.As long as the pressure of the fluid does not exceed the contact stress of the o-ring, leakage should not occur.Figure 1 - O-Ring Groove Dimensions At zero gauge pressure, only the pre-compressed resiliency of the o-ring provides the seal (see Figure 2). If the system pressure is in the range of 0 - 100 or even to 400 pounds per square inch (psi) it can be considered as low pressure (in this paper 400 psi or less is considered low pressure), and the seal is maintained predominantly by the pre-compression or squeeze on the o-ring and its resulting contact stress. As the pressure increases, the o-ring is forced to the low pressure side of the gland. This provides an additional increase in contact stress as the o-ring deforms to a D shape (see Figure 3) and the contact area of sealing under pressure increases to almost twice the original zero-pressure area. For this reason, an o-ring can easily seal a high pressure as long as it does not mechanically fail.3 of 7

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 41 of 44 Figure 2 - O-ring un-pressurized Figure 3 - O-ring with high pressure Best Design Practices:

a. The flexible nature of o-ring materials accommodates imperfections and/or waviness in the gland parts. But it is still important to maintain a good surface finish of those mating parts. The following best practices are suggested: 32 micro-inch finish on the contact surfaces (top of gland and bottom or groove); 63 micro-inch finish on the sides of the groove; machined radii in bottom of groove of 1/32 (reference 2); holding waviness of groove bottom to less than 2% of o-ring thickness per 12 length of groove.

Figure 4 shows a poor surface finish and affect the tool mark direction. Such a finish can cause leaks.Figure 4 - Surface Finish Can Cause Leaks

~~b. As for o-ring compression or squeeze it is a result of three factors: the force to compress the o-ring, durometer, and cross section thickness. A 15-20%~~

compression for dynamic (moving) applications (to mitigate wear) and 35-40% for static applications (reference 3) is generally suggested. Whereas, reference 2 (Parker) recommends 16% for dynamic applications and 30% for static.However, analysis and testing of the application will determine the ultimate compression. Compression% is defined as the deflection of the seal divided by the cross-section thickness (cord diameter) and the results times 100.4 of 7

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 42 of 44

c. The rigidity of the gland closure and closure bolting spacing must be adequate to compress the o-ring without deflecting. Any deflection will reduce the design compression of the seal.
d. The area of the cross-section of the gland should be in the range of 15 to 40%

greater than the area of the cross-section of the o-ring. A 75% fill is suggested which leaves 25% empty space in the gland groove (reference 2). However, it is very important that the installed o-ring contact the low pressure side of the gland groove (see Figure 1 in Appendix) such that the o-ring only has to move very little when pressurized or activated.

e. For o-ring gland grooves that are non-circular, the groove turn radii in rectangular and square layout (i.e. the corners), must be large enough so the o-ring will not kink and such that the o-ring will fully contact the low pressure side of the groove (Figure 5 shows an example of too sharp of a corner in the o-ring groove).

Otherwise, the small radius turn may impede any activation of the seal. For example, the radius in the corners of an o-ring groove layout is suggested to be at least 2 inches for a 1/4 (0.275) nominal diameter o-ring stock, or 7 to 8 times the o-ring diameter. Fabricate the o-ring to snugly fit the low pressure side of the gland groove all the way around including the corner radii.Figure 5 - Too Sharp of Corner

f. It is important not to stretch the o-ring since stretch affects seal compression by reducing cross section, which reduces the sealing potential of the o-ring. A stretch greater than 5% on the o-ring I.D. (equivalent ID in non-circular case) is not recommended because it can lead to a loss of seal compression (reference 3).
g. When sizing an o-ring, choose the largest cross-section thickness as practical.

The larger the cross-section, the more effective the sealing and longer the life of the seal. However, with a dynamic application in which friction is a factor, a compromise will be required.

h. A 70 durometer (shore A) hardness should be used in the design whenever possible since it usually has the best combination of properties for most applications. It provides good conformability versus a mid-range contact stress capability (see Graph 1). It is also considered the standard o-ring hardness and is readily available from suppliers.
i. A lubricant compatible with the o-ring material should be applied to the o-ring as it is installed to decrease friction and assist in the activation of the seal under pressure.

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Appendix A Document 3960C, Rev. 0, Attachment 3 Page 43 of 44

j. O-ring re-use: reusing an o-ring in an assembly after some time in service is generally not recommended. Is the o-ring deformed, cracked, harder than when new, discolored, or less than clean? When in doubt, change out.

2.0 ANALYSIS OF O-RING SEAL DESIGNS The maximum sealing capability of high pressure o-ring seal designs is dependent on seal activation as discussed earlier and is not solely dependent on the initial contact pressure as determined by the compression of the seal in its housing (groove).However for low pressure designs, an analysis of the contact stress makes it possible to better predict the success of the seal while assuming no activation of seal. Two methods are presented below:Parker Method:As an example referring to the Parker O-ring Handbook, Figures 2-8 (reference 2) ,below is an analysis of a 1/4 o-ring design, Shore A durometer, with 20% compression:For a 20% compression on a 1/4 inch nominal o-ring, , the compression load per linear inch of the seal is at around 35 pounds from the Parker Figures. Referring to the paper O-rings for Low Pressure Service, (reference 1), the contact area b per linear inch of seal is estimated by b=2.4x, where x is the deflection of the o-ring cross-section.Therefore, the contact area per inch is (2.4) (.20)(.275) = .132 square inch. The contact stress Smax per linear inch just from pre-compressed resiliency of the seal is 35 pounds divided by .132 square inches yielding 265 psi. This means, in theory with all things being perfect, the seal should not leak until the pressure exceeds 265 psig, as a minimum.Lindly Method:As an example using Lindlys analysis as referred to in reference 4, below is an analysis of a 1/4 o-ring design, Shore A durometer, with 20% compression:Lindly derived the equation below for predicting the contact stress Smax with respect to the modulus of elasticity E (E can be derived from Shore A durometer):Smax=E (0.849(1.25G1.5 + 50G6))0.5 where: Smax = contact stress psi, G = fractional compression, (20% = 0.2), E= modulus of elasticity (which can be found in reference 1, of E versus Shore A durometer; 60 = 630 psi, 70= 1,040 psi, 80 = 1,705 psi)Smax is plotted in Graph 1 for Shore A durometer 60, 70, and 80. The Lindly Method gives a higher contact stress than the Parker, however, due to the simplicity of use, the Lindly Method is the preferred method.Calculating contact stress is only a starting point in the seal design for determining the minimum required compression, however, with variables in the less than perfect 6 of 7

Appendix A Document 3960C, Rev. 0, Attachment 3 Page 44 of 44 sealing system, a margin above that is prudent, so it is suggested to follow the best practices in Section 1.0 in order to achieve a successful seal.Graph 1 - Contact Stress vs. Compression (Lindly) 1100 1000 900 Durometer - Shore A 800 60 70 80 Contact Stress- PSI 700 600 500 400 Low Pressure Zone 300 200 100 05 10 20 30 40

~~% Compression 3.0 APPENDIX~~
~~References:~~
1. Hertz,D.L.,1979, "O-Rings for Low Pressure Service", Machine Design, 4/12/79, pp.94-98 (note, paper applies mainly to dynamic applications)
2. Parker Hannifin Corporation, O-Ring Handbook, Catalog ORD 5700A/US
3. Apple Rubber Products Inc.,www.applerubber.com
4. Green, Itzhak and English, Capel, Stresses and Deformation of Compressed Elastomeric O-Ring Seals NOTE DISCLAIMER: The information and calculations are provided herein "as is" without any express or implied warranties. While effort has been taken to ensure the accuracy of the information and calculations, the authors/maintainers/contributors assume no responsibility for errors or omissions, or for damages resulting from their use. The contents of the information or calculations herein might be totally inaccurate, inappropriate, or misguided. There is no guarantee as to the suitability of said information or calculations for any purpose. Use at your own risk.

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Document 3960C, Rev. 0, Attachment 4 Page 2 Revisions Rev. DCR/N Pages Description of Changes No. No. Affected 0 N/A Initial release All Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 3 Table of Contents Page 1 OBJECTIVE AND SCOPE 5 1.1 OBJECTIVE 5 1.2 SCOPE 5 1.3 HISTORICAL LEAKAGE 8 2 METHODOLOGY 10 2.1 VARIABLES 10 2.2 TOTAL SEALING LOAD AND CONTACT STRESS 11 2.2.1 Calculation Approach for Piston Check Valve with Soft Seat 12 2.2.2 Calculation Approach for Piston Check Valve with Metal Seat 15 2.2.3 Calculation Approach for Inline Check Valve 18 3 INPUTS 19 3.1 CALCULATION INPUTS 19 3.1.1 LLTR Test Pressure and Adjusted Maximum Containment Design Pressure 19 3.1.2 Mean Seat Contact Diameter and Minimum Valve Cracking Pressure 19 3.1.3 Seat Contact Width for Piston Check Valves with Soft Seat 20 3.1.4 Design Input for Inline Check Valve Group 43-1 20 3.1.5 Design Input for Metal-Seated Check Valves Group 62-4 21 4 ASSUMPTIONS 22 5 RESULT, CONCLUSION AND RECOMMENDATION 25 5.1 SOFT SEAT PISTON CHECK VALVES 25 5.2 METAL SEATED PISTON CHECK VALVES (GROUP 62-4) 26 5.3 INLINE CHECK VALVE (GROUP 43-1) 26 5.4 RECOMMENDATION 27 6 REFERENCES 28 Appendix A - Supporting Documents Description Pages Main Text 29 Appendix A 66 Total Pages 95 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 4 List of Tables Table Description Page Table 1-1: Analysis Scope 6 Table 1-2: Leakage History of Valves in Unit 1 and Unit 2 8 Table 3-1: Common Input Data 19 Table 3-2: Mean Seat Contact Diameter, Minimum Valve Cracking Pressure, and Seat Contact Width 20 Table 5-1: Seat Load and Contact Stress Results for Soft Seat Piston Check Valve 25 Table 5-2: Percentage Increase in Leakage Flow Area Calculation Results 26 Table 5-3: Peak Seat Contact Stress For Material With 60, 70 And 80 Durometer Shore A Hardness 27 List of Figures Table Description Page Figure 2-1: Inline Check Valve [3c] 11 Figure 2-2: Piston Check Valve [13] 12 Figure 2-3: Piston Check Valve with Soft Seat Insert with Screw [3a] and Resilient Seated Disc[3d] 13 Figure 2-4: Mating of Seat Joint Surface 16 Figure 2-5: Surface Asperities (High Spot) on a Seat Contact Band 17 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 5 1OBJECTIVE AND SCOPE 1.1 OBJECTIVE Kalsi Engineering, Inc. (KEI) has been contracted by Tennessee Valley Authority (TVA) to provide engineering services to evaluate the impact of local leak rate test (LLRT) pressures higher than the calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA), Pa, for cases where higher test pressure tends to increase the sealing force. This work is being done in accordance with the scope defined in Purchase Order No. 6232543 [2] 1.The objective of this report is to determine the impact of the reduced LLRT pressure from DPtest to Pa on the seat leakage. All work performed under this project was done in accordance with the requirements of the KEI Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B requirements.1.2 SCOPE The scope of this attachment is LLRT lift and piston check valves. Valve IDs and basic information are shown below in Table 1-1. Group 61-1 and 67-1 have the same check valve as can be seen from the drawing number in Table 1-1. Hence, they will be treated identical for this assessment.Group 31-1 and Group 67-3 have identical disc, seat and spring part numbers as seen from the drawings [3a] and [3g] respectively. They will have similar seat sealing characteristics and are treated as identical valve for this assessment. Drawing for Group 32-2 [3b] shows metal seated configuration and an alternate soft seat assembly. All the valves in the Group 32-2 (both Unit 1 and Unit 2) have soft seat assembly per TVA documentation [12].1 The number in [] indicates reference number documented in Section 6.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 6 Table 1-1: Analysis Scope Seat Vendor Dwg.Group Component ID Comp Description Manufacturer Type No.INCORE INSTR RM 2-CKV-31-3378 AHU 2B CWS LEAK RATE CHECK INCORE INSTR RM 2-CKV-31-3392 AHU 2B CWR LEAK TVSW- RATE CHECK F990/FLOWSERVE 31-1 SOFT 30604GS-(2) INCORE INSTR RM CORPORATION 2-CKV-31-3407 AHU 2A CWS LEAK RATE CHECK INCORE INSTR RM 2-CKV-31-3421 AHU 2A CWR LEAK RATE CHECK CONTROL AIR 1-CKV-32-293 CNTMT CHECK 1-CKV-32-303 ESSENT CNTL AIR 1-CKV-32-313 CNTMT CHECK TVD-D-9911- ESSENT CNTL AIR 32-2 SOFT 2 2-CKV-32-323 K085/KEROTEST (2) CNTMT CHECK ESSENT CNTL AIR 2-CKV-32-333 CNTMT CHECK CONTROL AIR 2-CKV-32-343 CNTMT CHECK 1-CKV-43-834 PAS WASTE TO CNTMT SUMP 1-CKV-43-841 CHECK 43-1 SOFT N89-180 1-CKV-43-883 PAS CIRCLE SEALS CONTAINMENT 1-CKV-43-884 AIR RETURN CHECK GLYCOL SUPPLY 1-CKV-61-533 HEADER BYPASS CHECK GLYCOL RETURN 1-CKV-61-680 HEADER BYPASS CHECK GLYCOL COOLED 1-CKV-61-692 FLOOR SUPPLY BYPASS CHECK Flowserve/A391/ANCHOR-61-1 SOFT W9825144 GLYCOL COOLED DARLING 1-CKV-61-745 FLOOR RETURN BYPASS CHECK GLYCOL SUPPLY 2-CKV-61-533 HEADER BYPASS CHECK GLYCOL RETURN 2-CKV-61-680 HEADER BYPASS CHECK 2See Reference [12]Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 7 Seat Vendor Dwg.Group Component ID Comp Description Manufacturer Type No.GLYCOL COOLED 2-CKV-61-692 FLOOR SUPPLY BYPASS CHECK GLYCOL COOLED 2-CKV-61-745 FLOOR RETURN BYPASS CHECK CVCS SEAL WTR 1-1-CKV-62-639 FCV-62-61 EQL TVD-D-9556- CHECK 62-4 HARD K085/KEROTEST (2) CVCS SEAL WTR 2-2-CKV-62-639 FCV-62-61 EQL CHECK 1-CKV-63-868 1-CKV-63-868 CONTAINMENT N2 HEADER CHECK K085/KEROTEST TVW1-63-1 SOFT MANUFACTURING 30608GS-(2) 2-CKV-63-868 CORP.2-CKV-63-868 CONTAINMENT N2 HEADER CHECK 1-FCV-67-87 1-CKV-67-575A BYPASS CHECK 1-FCV-67-103 1-CKV-67-575B BYPASS CHECK 1-FCV-67-95 1-CKV-67-575C BYPASS CHECK 1-FCV-67-111 1-CKV-67-575D BYPASS CHECK 67-1 SOFT W9825144 ANCHOR-DARLING 2-FCV-67-87 2-CKV-67-575A BYPASS CHECK 2-FCV-67-103 2-CKV-67-575B BYPASS CHECK 2-FCV-67-95 2-CKV-67-575C BYPASS CHECK 2-FCV-67-111 2-CKV-67-575D BYPASS CHECK 1-FCV-67-295 1-CKV-67-585A BYPASS CHECK 1-FCV-67-297 1-CKV-67-585B BYPASS CHECK 1-FCV-67-296 1-CKV-67-585C BYPASS CHECK 1-FCV-67-298 1-CKV-67-585D BYPASS CHECK 67-3 SOFT 72576978 FLOWSERVE 2-FCV-67-295 2-CKV-67-585A CORPORATION BYPASS CHECK 2-FCV-67-297 2-CKV-67-585B BYPASS CHECK 2-FCV-67-296 2-CKV-67-585C BYPASS CHECK 2-FCV-67-298 2-CKV-67-585D BYPASS CHECK Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 8 Seat Vendor Dwg.Group Component ID Comp Description Manufacturer Type No.PRESSURIZER 1-CKV-68-849 RELIEF TANK N2 TVA- SUP HDR CHECK 68-1 SOFT K085/KEROTEST 30508GLS-(2) PRESSURIZER 2-CKV-68-849 RELIEF TANK N2 SUP HDR CHECK 1.3 HISTORICAL LEAKAGE Reference 4 and Reference 15 provide the LLRT history for the Unit 1 and Unit 2 check valves respectively. Table 1-2 summarizes the results of the tests.Table 1-2: Leakage History of Valves in Unit 1 and Unit 2 Seat Leakage Group Vendor Dwg. No. Component ID Type history 2-CKV-31-3378 Unfavorable TVSW-30604GS- 2-CKV-31-3392 Unfavorable 31-1 SOFT (2) 2-CKV-31-3407 Favorable 2-CKV-31-3421 Favorable 1-CKV-32-293 Favorable 1-CKV-32-303 Favorable 1-CKV-32-313 Favorable 32-2 SOFT TVD-D-9911-(2) 2-CKV-32-323 Favorable 2-CKV-32-333 Favorable 2-CKV-32-343 Favorable 1-CKV-43-834 Favorable 1-CKV-43-841 Favorable 43-1 SOFT N89-180 1-CKV-43-883 Favorable 1-CKV-43-884 Favorable 1-CKV-61-533 Favorable 1-CKV-61-680 Favorable 1-CKV-61-692 Favorable 1-CKV-61-745 Favorable 61-1 SOFT W9825144 2-CKV-61-533 Favorable 2-CKV-61-680 Favorable 2-CKV-61-692 Favorable 2-CKV-61-745 Favorable 1-CKV-62-639 Unfavorable 62-4 HARD TVD-D-9556-(2) 2-CKV-62-639 Favorable TVW1-30608GS- 1-CKV-63-868 Favorable 63-1 SOFT (2) 2-CKV-63-868 Favorable 67-1 SOFT W9825144 1-CKV-67-575A Favorable Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 9 Seat Leakage Group Vendor Dwg. No. Component ID Type history 1-CKV-67-575B Favorable 1-CKV-67-575C Favorable 1-CKV-67-575D Unfavorable 2-CKV-67-575A Favorable 2-CKV-67-575B Favorable 2-CKV-67-575C Favorable 2-CKV-67-575D Favorable 1-CKV-67-585A Favorable 1-CKV-67-585B Favorable 1-CKV-67-585C Favorable 1-CKV-67-585D Favorable 67-3 SOFT 72576978 2-CKV-67-585A Favorable 2-CKV-67-585B Favorable 2-CKV-67-585C Favorable 2-CKV-67-585D Favorable TVA-30508GLS- 1-CKV-68-849 Favorable 68-1 SOFT (2) 2-CKV-68-849 Favorable Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 10 2METHODOLOGY 2.1 VARIABLES Variable Description Units Ao Area based on mean seat diameter = 4 2 in2 As Nominal seat contact area for metal seated check valve in2 Ahs Cross-sectional area of asperities (high spots) in2 Percentage increase in leakage flow area when pressure reduces from AI %DPtest to Pa Percentage area of the nominal seat contact area due to asperity Aasp %contacts (actual seat contact area)CL Seat leakage coefficient proportionality constant dm Mean seat contact diameter in dr Maximum Cross section diameter of O-ring in da Diameter of surface asperities in DP Difference between upstream and downstream pressure psi DPtest Bounding LLRT test differential pressure psig Es Elastic modulus of O-ring material psi E Elastic modulus of disc/seat material psi FS Sealing load due to differential pressure lbf FS_DP Sealing load due to differential pressure at DPtest lbf FS_Pa Sealing load due to differential pressure at Pa lbf FS_Total Total sealing load at DPtest lbf Fspring Sealing load due to spring lbf FS_Total_red Reduced total sealing load at Pa lbf Fhs Seat load per asperity (high spot) lbf ha Height of the asperities (high spots) at no pressure/seat load in ha_DPtest Height of the asperities (high spots) at DPtest in ha_Pa Height of the asperities (high spots) at Pa in ha Change in the height of asperities (high spots) in Nhs Number of asperities (high spots) on the seat contact band Calculated peak containment internal pressure related to the design-Pa psig basis loss-of-coolant accident (LOCA)Pcr Check valve cracking pressure psi Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 11 Smax Peak seat contact stress psi t Seat contact band width of lift check valve in R Percentage reduction in sealing load due to Pa vs. test DP %RP Percentage reduction of sealing load due to differential pressure %x O-ring compressive displacement in Average seat contact stresses psi DPtest Average seat contact stress at DPtest psi Pa Average seat contact stress at Pa psi O-ring normalized squeeze 2.2 TOTAL SEALING LOAD AND CONTACT STRESS The lift check valves also known as globe check valves are analyzed in this attachment. They rely on the spring force, weight, and differential pressure of the fluid to provide seal.During LLRT, sealing load includes differential pressure force, FS, spring force, Fspring, and force due to the disc weight. All three force components are assisting the sealing action. Differential pressure load is proportional to the DP and the area over which the DP acts. The spring force is calculated using the cracking pressure of the valve. The weight of the disc assembly will be very small compared to the DP force and is excluded from the total sealing load which is conservative.This assessment includes two types of lift check valves, 1) inline check valve (Figure 2-1) and 2) piston check valve (Figure 2-2).Figure 2-1: Inline Check Valve [3c]Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 12 Figure 2-2: Piston Check Valve [13]2.2.1 Calculation Approach for Piston Check Valve with Soft Seat The piston check valve in Groups 32-2, 61-1, 67-1, 31-1, 67-3, 63-1 and 68-1 have soft seated elastomeric insert installed on the disc. The soft seat on these valves is installed in two different ways on the disc as shown in Figure 2-3.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 13 Figure 2-3: Piston Check Valve with Soft Seat Insert with Screw [3a] and Resilient Seated Disc [3d]Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 14 Initial contact develops between the seat and the elastomeric insert as the disc contacts the seat.The contact between the elastomeric insert and the seat will develop contact stresses. The contact stresses higher than the fluid differential pressure will ensure a positive sealing margin. The following calculations determines the seat contact stress for these valves.The spring force acting on the seat assist the sealing load and is calculated using the check valve cracking pressure and the mean seat contact diameter.

~~= (1)~~
~~Where, 2~~
=

4 The differential pressure acts normal to the valve disk surface and produces a sealing force equal to the differential pressure (DP) times the area over which the DP acts. The area defined by the seat contact diameter is used for the area over which the DP acts. The DP induced sealing force is given by Equation 2.

~~= (2)~~

Total seat load is equal to the sum of the spring force and the force due to differential pressure acting on the seat._ = + (3)Under zero pressure condition, the soft insert and the metal seat develops a contact due to the spring force. As the differential pressure is applied across the valve, the contact width between the soft insert and the metal seat grows. The seat contact width is used for calculating the seat contact area. A larger seat contact width will develop lower average seat contact stresses. The average seat contact stress is calculated using Equation 4._ (4)

It is assumed that the seat contact width will stay constant when the differential pressure changes from 16.5 psi to 9 psi. This is a conservative assumption because any decrease in the seat contact width will increase the seat contact stress per Equation 4. Therefore, this assumption does not require a verification.The percentage reduction in total sealing load, R, due to reduction in pressure from DPtest to Pa is determined using Equation 5._ _ (5)

~~= .100 Non-Proprietary Version~~

Document 3960C, Rev. 0, Attachment 4 Page 15 2.2.2 Calculation Approach for Piston Check Valve with Metal Seat The piston check valves in Group 62-4 have metal seated disc. The sealing load in a metal-seated piston check valve, without any spring load or disc weight, is equal to the differential pressure force, FS acting on the disc. FS is equal to the DP times the area over which the DP acts as shown in Equation 2.The area over which DP acts, Ao, is function of the geometry of the disc and seat and remains the same for all DP. Therefore, the percentage reduction in the sealing load, RP, when the differential pressure reduces from DPtest to Pa is equal to the percentage reduction in the pressure which is given by:16.5 9.0 (6)

~~= 100 = 100 = 45.5%~~

16.5 Unlike the soft-seated piston check valves, a tight sealing of a metal-seated valve requires yielding of one material into the surface waviness and surface roughness of the other to block direct leakage paths. Even, seemingly smooth machined surfaces have surface asperities as illustrated in Figure 2-4. When the two surfaces contact each other, the surface asperities initially establish the contact. With an increasing load, the asperities initially deform elastically and then plastically. To ensure a reliable seal, the surface asperities within the contact band need to deform plastically over a reasonable amount of bandwidth. At low pressures, the seat load will not be sufficient to plastically yield the asperities on the contacting surfaces and therefore, the asperities will deform elastically. Due to the fact that these valves were providing a reliable sealing at DPtest pressure, the leak path at this pressure will have a very high flow resistance. Any reduction in the seat load will decrease the flow resistance by increasing the leakage flow area, AI, due to a reduced elastic deformation of the asperities (high spots). As shown in Equation 6, the seat load will decrease by 45.5% when the pressure reduces from the DPtest of 16.5 psig to Pa of 9.0 psig. The reduction in the seat load will reduce the leak path flow resistance. A reduction in the flow resistance is equivalent to an increase in the leakage coefficient, CL. Based on Equation 3-3 in Section 3.2 of the main report, the CL can increase by 35% before the measured leakage at pressure Pa would increase from the measured leakage at DPtest.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 16 Figure 2-4: Mating of Seat Joint Surface A calculation has been performed to estimate an increase in the leakage flow area, AI, when the pressure is reduced from DPtest to Pa. The calculation is based on a simplified but conservative assumption (Assumptions 6 to 10 in Section 4) of a leak flow area developed by surface asperities (high spots) on the disc/seat surfaces that comes in contact when the piston check valve disc closes.The idea behind this calculation is to determine the effect of the seat load change on the height of the asperities that in turns increases or decreases the leakage flow area. Figure 2-5 shows asperities of an exaggerated size on a seat contact band in the top view. For better visualization, only one high spot is shown in the side view. The side view shows a leakage flow channel developed by a high spot at zero seat load, at DPtest, and at Pa. The flow channel width is constant, whereas the height of the asperities, ha, changes as it gets compressed due to seat load.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 17 Figure 2-5: Surface Asperities (High Spot) on a Seat Contact Band For better visualization, only one asperity is shown in the side view. The side view shows the flow channel at no pressure, at DPtest, and at Pa. The flow channel width is constant, whereas the height, ha, changes as the seat load compresses the surface asperities.The percentage increase in leakage flow area, AI, is calculated using Equation 7. The AI equation does not account for a change in diameter of the asperities due to a lateral strain while under compression. The lateral strain of the asperities will change the effective width of the flow channel.The change in effective width of the flow channel will be orders of magnitude smaller than the flow channel width and will have a negligible effect on AI._ _ (7)

~~= x 100~~
~~Where,~~

_ = _ (8)_ = _ (9)The change in height of asperities due to the seat load is calculated using Equation 10:Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 18 (10)

=
~~Where,~~
=

(11) 2

4 (12) 100

~~= (13) = (14)~~

Table 5-2 documents the percentage increase in the leakage flow area, AI which is based on the design inputs in Section 3.1.5 and Assumptions 6 to 10 in Section 4.2.2.3 Calculation Approach for Inline Check Valve The inline check valve in Group 43-1 is shown in Figure 2-1. Figure 2-1 [3c] shows that an O-ring (Item 5) provides the primary seal at low pressures. This O-ring is installed on the poppet (Item 4). When the valve closes, the O-ring (Item 5) gets squeezed between the poppet (Item 4) and end (Item 8) providing a primary seal at low pressures. Theoretically, the peak seat contact stress, Smax, higher than the fluid differential pressure will ensure a positive sealing margin. The peak seat contact stress is calculated using Equation 15 [8] which depend on the amount of O-ring compressive displacement, x, and the modulus of elasticity, Es.1 16 1.5 6 2

~~= 6 (1.25 + 50 ) (15)~~

In the above equation, Es is the elastic modulus and is a function of the Shore A hardness of the O-ring material. The normalized squeeze, , which is a ratio of compressive displacement, x, and the O-ring cross section diameter, dr, is given by Equation 16 [8].

~~= (16)~~

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 19 3INPUTS 3.1 CALCULATION INPUTS The input data for the analyses are documented in this section and were obtained from calculations, drawings, specifications, and other information provided by TVA. Inputs that require additional clarification are documented below, if applicable. Justified assumptions were made where data were not available. It is important to note that the results of this analysis may be significantly affected by changing key inputs. It will be necessary to perform an impact analysis if key data are changed in the future.Table 3-1: Common Input Data Item Variable Value Reference Area based on mean seat diameter Ao Calculated LLRT test differential pressure, psi DPtest 16.5 See 3.1.1 Calculated peak containment internal pressure Pa 9.0 See 3.1.1 related to the design-basis loss-of-coolant accident (LOCA) 3.1.1 LLTR Test Pressure and Adjusted Maximum Containment Design Pressure The maximum permissible LLRT test pressure is 1.1 x 15= 16.5 psig [9, 14]. Pa for Watts Bar is 9.36 psig [9]. For purposes of this analysis, a lower and more conservative value of 9 psig is used.3.1.2 Mean Seat Contact Diameter and Minimum Valve Cracking Pressure The mean seat contact diameter are provided by TVA [11] which are documented in Table 3-2.The mean seat diameter for Groups 31-1 and 67-3 is not available. The mean seat diameter is assumed to be equal to the valve flow bore diameter which is 0.625 inch [3a]. The valve flow bore diameter looks larger than the mean seat diameter [3a]. A larger mean seat diameter will develop lower average seat contact stresses which is conservative. Therefore, this assumption does not require a verification.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 20 The cracking pressure are provided by TVA [11] which are documented in Table 3-2. The minimum cracking pressure for Groups 32-2, 67-3, and 31-1 are not available. For conservatism, the minimum cracking pressure for these valves is set to zero. This means the spring force for these valves is not included in the average seat contact stress calculation. The check valve in Group 68-1 does not have a spring in the assembly as shown in the Reference 3h, therefore, the valve cracking pressure is set to zero.3.1.3 Seat Contact Width for Piston Check Valves with Soft Seat The seat contact width, between the elastomeric insert of the disc and the metal seat, for the valve Groups 61-1 and 67-1 is 0.06 inches based on the scaling from Reference 3e which is equal to 6%of the mean seat diameter. The scaling of seat contact width is not possible for the valves in Groups 31-1 and 67-3 [3a], 63-1 [3f], 32-2 [3b], and 68-1 [3h]. The seat contact width for the valves in these groups is assumed to be equal to 10% of the mean seat contact diameter but no larger than 1/8 inches. The seat contact width does not vary proportionally with the mean seat contact diameter. Typically, the seat contact width is smaller than 0.10 inches. A larger seat contact width will provide lower average seat contact stresses. Therefore, the assumption of upper bound seat contact width of 0.125 inches is conservative and it does not require verification.The mean seat contact diameter, minimum valve cracking pressure, and seat contact width are summarized in Table 3-2.Table 3-2: Mean Seat Contact Diameter, Minimum Valve Cracking Pressure, and Seat Contact Width Mean Seat Minimum Valve Seat contact Group Contact Diameter Cracking width (t), in (dm), in Pressure (Pcr), psi 32-2 2.500 0 0.125 61-1/67-1 1.000 5 0.060 63-1 0.750 3 2 0.075 67-3/31-1 0.625 0 0.063 68-1 1.375 0 0.125 3.1.4 Design Input for Inline Check Valve Group 43-1 The O-ring part number used in this inline check valve per Reference 3d is AS568-211. The cross-section diameter for the O-ring size 211 is 0.139+/-.004 inch [6]. The durometer hardness of the O-3 Table in Reference 11 shows 1.0 for dm and email dated 07/31/2020 states this is equal to 0.75. Therefore 0.75 is considered as mean seat contact diameter for conservatism.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 21 ring is not known but typically an O-ring with a Shore A Hardness of 70-durometer that has approximate room temperature elastic modulus of 1040 psi [7] is used in such applications. A softer O-ring with a lower elastic modulus will provide a conservative peak seat contact stress; therefore, the O-ring Shore A Hardness of 60-durometer with the elastic modulus of 630 psi [7] is used in this analysis.3.1.5 Design Input for Metal-Seated Check Valves Group 62-4 The mean seat contact diameter for the group 62-4 per Reference [11] is equal to 1.375 inches.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 22 4ASSUMPTIONS Data that have not been formally verified are treated as assumptions. Where possible, the basis of the data has been noted. The following general assumptions were used in this analysis.

1. It is assumed that the seat contact width will stay constant when the differential pressure changes from 16.5 psi to 9 psi. This is a conservative assumption because any decrease in the seat contact width will increase the seat contact stress per Equation 4. Therefore, this assumption does not require a verification.
2. The seat contact width for the valves in Groups 31-1, 67-3, 63-1, 32-2, and 68-1 is assumed to be equal to 10% of the mean seat contact diameter but no larger than 1/8 inches (see Table 3-2). The seat contact width does not vary proportionally with the mean seat contact diameter. Typically, the seat contact width is smaller than 0.10 inches. A larger seat contact width will provide lower average seat contact stresses. Therefore, the assumption of upper bound seat contact width of 0.125 inches is conservative and it does not require verification.
3. The disc weight induced seat load component will be small compared to the DP induced seat load and is excluded from the total seat load calculation. Not including the disc weight induced seat load in the total seat load calculation will increase the percentage reduction in the total seat load when pressure reduced from 16.5 psi to 9.0 psi. Therefore, this is a conservative assumption and does not require verification.
4. The minimum cracking pressure of 0 psi is assumed for valves in Groups 32-2, 67-3, and 31-1. Not including the spring force induced seat load in the total seat load calculation will increase the percentage reduction in the total seat load when pressure reduced from 16.5 psi to 9.0 psi. Therefore, this is a conservative assumption and does not require verification.
5. The mean seat contact diameter for Groups 31-1 and 67-3 is not available. The mean seat contact diameter is assumed to be equal to the valve flow bore diameter which is 0.625 inch [3a]. The valve flow bore diameter looks larger than the mean seat diameter [3a]. A larger mean seat diameter will develop lower average seat contact stresses which is conservative. Therefore, this assumption does not require a verification.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 23 The following assumptions are made to calculate percentage change in leakage flow area discussed in Section 2.2.2.

6. The surface asperities (high spots) are assumed to be of a cylindrical profile. This is a simple geometry. Typically, the asperities have a profile with a smaller cross-sectional area at the contacting surfaces and the area increase towards the base of the asperities. Such profile will be much stiffer compared to a cylindrical profile. Therefore, the cylindrical profile will undergo a larger deformation due to the differential pressure force compared to the profile mentioned above. The larger deformation will result into a larger change in the leakage flow area for the which is conservative. Therefore, this assumption does not require a verification.
7. The height, ha, and diameter, da of the surface roughness high spots are assumed to be equal to 16 µin. This assumption is based on a surface roughness of 16 Ra. Typically, the sealing surfaces are machined to the lowest surface roughness to ensure tight sealing. A sensitivity analysis showed that this assumption has no effect on the calculated percentage increase in the leakage area.
8. The seat contact band width, t, for Group 62-4 valve is assumed to be of 0.005 inches which is a reasonable assumption. A lower seat contact band width will provide a conservative result. A sensitivity analysis showed that lowering the seat contact band width to 0.001 inches does not change the overall conclusion because the percentage increase in the leakage area, AI, remains below 0.1%. Therefore, this assumption does not require a verification.
9. It is assumed that the area covered with the surface asperities, Aasp, is 10% of the nominal seat contact area, As. Typically, the area covered with the asperities is higher than 10%. The seat load, FDP, acts on the area covered with the asperities therefore, lower the area, higher will be the deformation of the asperities which is conservative. A sensitivity analysis showed that lowering this value to 1% does not change the overall conclusion because the percentage increase in the leakage area remains negligible. Therefore, this assumption does not require a verification.
10. The modulus of elasticity, E, of the seat/disc material is assumed to be equal to 3.0e7 psi.

The valve drawings [3e] shows that the disc and seat surfaces are hard faced which typically has a higher modulus than the one used here. The lower modulus provides a conservative result. Therefore, this assumption does not require a verification.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 24 The following assumptions are made to calculate peak seat contact stress for Group 43-1 valves as discussed in Section 2.2.3.

11. The O-ring compressive displacement, x, (squeeze) is not known. Per Parker O-ring Handbook [6], a minimum squeeze of 0.007 inches is recommended for any cross-section O-ring. Therefore, the minimum recommended squeeze of 0.007 inches is assumed for the peak seat contact stress calculation in Equation 15. Table 1-2 shows that all the valves in the Group 43-1 have a favorable history of LLRT leakage results. So, we can assume that the O-ring used in this check valve satisfy the requirement of minimum recommended squeeze of 0.007 inches. Therefore, the value of 0.007 inches for the O-ring compressive displacement is a conservative assumption and does not require verification.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 25 5RESULT, CONCLUSION AND RECOMMENDATION This section documents the calculation results and conclusions derived based on the calculation approaches documented in Section 2, design inputs documented in Section 3, and the assumptions in Section 4.5.1 SOFT SEAT PISTON CHECK VALVES The average seat contact stress calculation results for the piston check valves in Groups 32-2, 61-1, 67-1, 31-1, 67-3, 63-1 and 68-1 are shown in Table 5-1.Table 5-1: Seat Load and Contact Stress Results for Soft Seat Piston Check Valve Valve FS_DP FS_Pa Fspring FS_Total FS_Total_red DPtest R (%) Pa(psi)Group (lbf) (lbf) (lbf) (lbf) (lbf) (psi) 32-2 81.0 44.2 0 81 44.2 45.5 82.5 45.0 61-1/13.0 7.1 3.9 16.9 11.0 34.9 89.6 58.3 67-1 63-1 7.3 4.0 0.9 8.2 4.9 40.5 46.3 27.5 67-3/5.1 2.8 0 5.1 2.8 45.5 41.3 22.5 31-1 68-1 24.5 13.4 0 24.5 13.4 45.5 45.4 24.8 The percentage reduction in the seat load, R, varies between 35% to 46%. However, the average seat contact stresses (DPtest and Pa) are higher than the differential pressures at both Pa and DPtest pressures. The peak seat contact stress is not calculated for these valves, but it will be higher than the average seat contact stresses calculated above. The peak seat contact stress must exceed the differential pressure to ensure sealing. Therefore, the leakage is not expected to increase when the pressure is reduced from DPtest to Pa for the valves in these groups.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 26 5.2 METAL SEATED PISTON CHECK VALVES (GROUP 62-4)The results of the calculation approach documented in Section 2.2.2 are included here. The calculation is performed based on the design inputs in Section 3.1.5 and Assumptions 6 to 10 in Section 4. Table 5-2 documents the percentage increase in the leakage flow area, AI.Table 5-2: Percentage Increase in Leakage Flow Area Calculation Results Valve Group 62-4 dm, in 1.375 As, in2 0.022 FS_DP, lb 25.40 FS_Pa, lb 13.85 Nhs 1.074E+07 Fhs_DPtest, lb 2.36E-06 Fhs_Pa, lb 1.29E-06 ha_DPtest, in 0.006 ha_Pa, in 0.003 AI,% 0.02 Table 5-2 shows that the calculated increase in the leakage flow area, AI, is 0.02% for Groups 62-4 valves which is negligible. Therefore, it not expected to increase the leakage coefficient, CL, by 35% which is a threshold for the measured leakage at the lower pressure, Pa, to increase from the measured leakage at the higher pressure, DPtest.5.3 INLINE CHECK VALVE (GROUP 43-1)The results of the calculation approach documented in Section 2.2.3 are included here. These results are based on the design inputs in Section 3.1.4 and Assumption 11 in Section 4.The normalized squeeze, , is calculated using Equation 16 and is equal to 0.05 based on the minimum squeeze of 0.007 inches and the maximum O-ring cross-section diameter of 0.143 inches (0.139+0.004).Table 5-3 documents the peak contact stresses calculated using Equation 15 for different O-ring normalized squeezes, and elastic modulus.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 27 Table 5-3: Peak Seat Contact Stress For Material With 60, 70 And 80 Durometer Shore A Hardness Durometer Hardness 60 (ES =630 70 (ES=1040 80 (ES=1705 psi) psi) psi)Peak Seat Contact Stresses (Smax), psi 0.05 68.6 113.3 185.7 0.1 115.5 190.6 312.5 0.15 157.0 259.2 425.0 0.2 196.9 325.0 532.8 0.25 238.3 393.3 644.8 0.3 285.5 471.3 772.6 0.35 343.8 567.5 930.4 0.4 419.0 691.7 1134.0 0.45 516.8 853.1 1398.6 0.5 642.0 1059.8 1737.5 The peak seat contact stress is 68.6 psi (see Table 5-3) for the normalized squeeze of 0.05 and the hardness of 60 durometer. The normalized squeeze for the minimum O-ring compressive displacement of 0.007 inches is 0.05. Typically, the minimum compressive displacement is achieved at a pressure well below 9.00 psig. Therefore, at 9 psig pressure, the peak seat contact stress will be well above 68.6 psi and which will ensure a positive sealing margin.Therefore, the sealing capability of the inline check valve (Group 43-1) will not be affected by the change in DP.5.4 RECOMMENDATION The calculations for metal seated check valve (Group 62-4) show that the leakage will not increase due to change in DP, but the leakage test results (Table 1-2) show a unfavorable leakage history for the valve. Therefore testing of the valves in this group is recommended.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 28 6REFERENCES

1. KEI Document No. 1500C Rev. 15; Kalsi Engineering, Inc. Quality Assurance Manual.
2. TVA Purchase Order 6232543, Rev. Num: 0.
3. Valve Drawings
a. Flowserve, Y-Type Check Valve, Order/Tag Information Size .50 Class 600, Dwg. No.

TVSW-30604GS Rev B, Approved Date: 08/18/2010.

b. Flowserve, 2 Series 1500 Y-Type Check Valve, Dwg. No. TVD-D-9911-(2) Rev A, Approved Date: 02/28/2011.
c. Circle Seal Corporation, Check Valve, Dwg. No. N89-180, Approved Date: 04/29/1976.
d. Flowserve, Piston Check Valve Socket Ends Stainless Steel with Resilient Seat and Non-Cobalt Trim Size:1/2 Class: 1878, Dwg. No. W9825144 Rev E, Approved Date:08/08/2012.
e. Flowserve, 3/4 Series 1500 Carbon Steel Y- Check Valve w/ Soft Seat, Dwg. No. TVD-D-9956-(2) Rev C, Approved Date:06/08/2010.
f. Kerotest, 1 Series 600 Y-type Check Valve, Dwg. No. TVW1-30608GS-(2), Approved Date: 12/02/1980.
g. Flowserve, Y- Check Valve Socket Weld Ends Size: .50 Class: 600, Dwg. No. TVW-D-30504-(2) Rev G, Approved Date: 09/15/2011.
h. Kerotest, 1 600# Y- Check Valve Stainless Steel w/ Soft Seat, Dwg. No. TVA-30508GLS-(2), Approved Date:7/24/1997.
4. TVA Engineering Work Request, EWR20MEC088032, Generate List of U1 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 06/09/20.
5. Circle Seal Control Check Valve Catalog, Corona, USA.
6. Parker Hannifin Corporation, O-ring Handbook, Catalog ORD 5700A/US.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 4 Page 29

7. Hertz, D.L.,1979, "O-rings for Low Pressure Service", Machine Design, 4/12/79, pp.94-98 (note, paper applies mainly to dynamic applications).
8. Green, Itzhak and English, Capel, Stresses and Deformation of Compressed Elastomeric O-ring Seals.
9. WBN UFSAR Section 6.2, Containment Systems.
10. O-ring Seal Design Best Practices, Rev 1, 2012.
11. Response from TVA For Valve Data Requested by KEI (7/16/2020).
12. TVA Clarification on Seat Type Used in The Lift Check Valves For Group 32-2 (08/17/2020).
13. Flowserve Instruction Manual for Piston Check Valves, Manual No 800-PC, March 2004.
14. ANSI/ANS-56.8-1994, American National Standard for Containment System Leakage Testing Requirements.
15. TVA Engineering Work Request, EWR20MEC026076, Generate List of U2 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 08/19/2020 Non-Proprietary Version

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 1 of 66 Appendix A SUPPORTING DOCUMENTS Page No.Title Page 1A Reference 3 2A Reference 5 10A Reference 6 11A Reference 7 13A Reference 8 20A Reference 10 33A Reference 11 40A Reference 12 42A Reference 13 53A Total Pages 66A

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A2 of A66 Document 3960C, Rev. 0, Attachment 4 0.06 in 1.01 in 0.63 in 0.62 in 0.6 2 in Group 31-1/67-3 Appendix A Page 2 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A3 of A66 Document 3960C, Rev. 0, Attachment 4 Appendix A Page 3 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A4 of A66 Group 43-1 Document 3960C, Rev. 0, Attachment 4 Appendix A Page 4 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A5 of A66 2.23 in 1 in 0.1 1.48 in Document 3960C, Rev. 0, Attachment 4 1.11 in 0.0 6 in 5 in 0.0 0.85 in 10 psi cracking 0.74 in pressure Approved(A)DSA 10/3/2015 INITIALS DATE 0.38 in .Project: WBN DISCIPLINE: MECHANICAL CONTRACT: 00403336 UNIT# 0 MAJOREQUIPDESCRIPTION:PistonCheckValve 5.47 in Appendix A System061,067 COMMENTS:Page 5 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A6 of A66 Document 3960C, Rev. 0, Attachment 4 Group 62-4 Appendix A Page 6 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A7 of A66 Document 3960C, Rev. 0, Attachment 4 Group 63-1 Appendix A Page 7 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A8 of A66 Document 3960C, Rev. 0, Attachment 4 Appendix A Page 8 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A9 of A66 Document 3960C, Rev. 0, Attachment 4 Appendix A Group 68-1 Page 9 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 10 of 66 200 Series 0 to 3000 psig Check Valves H200 Series 0 to 6000 psig Check Valves Features & Benefits Technical Data Quick opening/positive closing Body Construction Materials Aluminum, brass, steel, 303 or 316 stainless steel

Provides a wide range of adaptability O-ring Materials Buna N, ethylene propylene, fluorosilicone, Kalrez, neoprene, PTFE, and Viton Large flow capacity Operating Pressure 200 Series: to 3000 psig (207 bar)
The patented sealing principle effects H200 Series: to 6000 psig (414 bar) complete leakproof closing under all Proof Pressure 1.5 times operating pressure pressure conditions Rated Burst Pressure 200 Series: 2.5: 1 H200 Series: 4: 1 Zero leakage Cracking Pressure 0.1 to 25 psig (0.007 to 1.72 bar)
Compact, easy installation. Efficient inline Temperature Range 320° F to +550° F (196° C to +288° C) check valves piston reduces size and weight Based on o-ring & body material, see How to Order Floating o-ring Connection Sizes 18 to 2
The streamlined poppet and full ports Note: Proper filtration is recommended to prevent damage to sealing surfaces.

offer minimum restriction to flow How it Works Open Closing Closed Full flow passages offer minimum Floating o-ring automatically establishes O-ring only seals. Full pressure load is restriction to flow. Spring is completely line contact with conical metal surfaces of carried by metal-to-metal seat. Increasing removed from flow path poppet and seat to cushion closing and pressure increases sealing efficiency; insure perfect sealing. metal seat prevents any possibility of deformation or extrusion of o-ring.Circle Seal Controls 2301 Wardlow Circle

Corona, CA 92880 Phone (951) 270-6200
Fax (951) 270-6201 www.circlesealcontrols.com

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 11 of 66 0ARKER / 2ING (ANDBOOK recovery when the squeeze is less than .1 mm (.005 inch).Compression Recovery of Three O-Ring The three curves, representing three nitrile compounds, show Compounds When Light Squeeze is Applied very clearly that a good compression set resistant compound 100 Recovery After can be distinguished from a poor one only when the applied Compression of squeeze exceeds .1 mm (.005 inches).Recovery 70 Hours at 100°C (212°F)O-Ring Applications 75 Most seal applications cannot tolerate a no or zero squeeze Recovery is Essentially condition. Exceptions include low-pressure air valves, for which Independent of the "oating pneumatic piston ring design is commonly used, 50 Sample Thickness and some rotary O-ring seal applications. See the Dynamic O-Percent of Original Delection Ring Sealing, Section V, and Tables A6-6 and A6-7 for more information on pneumatic and rotary O-ring seal design.25

'LAND &ILL 0 The percentage of gland volume that an O-ring cross-section mm 0 0.1 0.3 0.4 0.5 displaces in its con"ning gland is called gland "ll. Most In. 0 0.005 0.010 0.015 0.020 Compression O-ring seal applications call for a gland "ll of between 60%

to 85% of the available volume with the optimum "ll being&IGURE #OMPRESSION 2ECOVERY OF 4HREE / RING 75% (or 25% void). The reason for the 60% to 85% range is

OMPOUNDS 7HEN ,IGHT 3QUEEZE IS !PPLIED because of potential tolerance stacking, O-ring volume swell and possible thermal expansion of the seal. It is essential to An assembled stretch greater than "ve percent is not recom- allow at least a 10% void in any elastomer sealing gland.

mended because the internal stress on the O-ring causes more rapid aging. Over "ve percent stretch may sometimes be used, however, if a shorter useful life is acceptable. / 2ING #OMPRESSION &ORCE The force required to compress each linear inch of an O-ring Of the commonly used O-ring seal elastomers, the reduc- seal depends principally on the shore hardness of the O-ring, tion in useful life is probably greatest with nitrile materials. its cross-section, and the amount of compression desired.Therefore, where high stretch is necessary, it is best to use Even if all these factors are the same, the compressive force ethylene propylene, "uorocarbon, polyurethane or neoprene, per linear inch for two rings will still vary if the rings are whichever material has the necessary resistance to the tem- made from different compounds or if their inside diameters peratures and "uids involved. are different. The anticipated load for a given installation is not "xed, but is a range of values. The values obtained from a 3QUEEZE large number of tests are expressed in the bar charts of Figures The tendency of an O-ring to attempt to return to its original 2-4 through 2-8 in Section II. If the hardness of the compound uncompressed shape when the cross-section is de"ected is the is known quite accurately, the table for O-ring compression basic reason why O-rings make such excellent seals. Obviously force, Table 2-3 may be used to determine which portion of then, squeeze is a major consideration in O-ring seal design. the bar is most likely to apply.In dynamic applications, the maximum recommended squeeze Increased service temperatures generally tend to soften is approximately 16%, due to friction and wear consider- elastomeric materials (at least at "rst). Yet the compression ations, though smaller cross-sections may be squeezed as force decreases very little except for the hardest compounds.much as 25%. For instance, the compression force for O-rings in compound N0674-70 decreased only 10% as the temperature was in-When used as a static seal, the maximum recommended creased from 24°C (75°F) to 126°C (258°F). In compound squeeze for most elastomers is 30%, though this amount may N0552-90 the compression force decrease was 22% through cause assembly problems in a radial squeeze seal design. In a the same temperature range.face seal situation, however, a 30% squeeze is often bene"cial because recovery is more complete in this range, and the seal Refer to Figure 3-6 for the following information:may function at a somewhat lower temperature. There is a The dotted line indicates the approximate linear change danger in squeezing much more than 30% since the extra in the cross section (W) of an O-ring when the gland stress induced may contribute to early seal deterioration. prevents any change in the I.D. with shrinkage, or the Somewhat higher squeeze may be used if the seal will not be O.D., with swell. Hence this curve indicates the change exposed to high temperatures nor to "uids that tend to attack in the effective squeeze on an O-ring due to shrinkage the elastomer and cause additional swell. or swell. Note that volumetric change may not be such a disadvantage as it appears at "rst glance. A volumetric The minimum squeeze for all seals, regardless of cross-sec- shrinkage of six percent results in only three percent tion should be about .2 mm (.007 inches). The reason is that with a very light squeeze almost all elastomers quickly take 100% compression set. Figure 3-5 illustrates this lack of 0ARKER (ANNIlN #ORPORATION s / 2ING $IVISION 3-9 2360 Palumbo Drive, Lexington, KY 40509 0HONE s &AX www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 12 of 66 0ARKER / 2ING (ANDBOOK

~~'UIDE FOR $ESIGN 4ABLE~~

)F $ESIRED $IMENSION 3ELECT #LOSEST 2EAD (ORIZONTALLY 4O $ETERMINE IS +NOWN FOR $IMENSION IN #OLUMN IN #OLUMN $IMENSION FOR B-1 Groove Dia. (male gland)Bore Dia. male gland A C Plug Dia. (male gland)G Groove width Static O-Ring Sealing A Bore Dia. (male gland)Plug Dia. male gland C B-1 Groove (male gland)G Groove width A-1 Groove Dia. (female gland)Tube OD female gland B D Throat Dia. (female gland)G Groove width A-1 Groove Dia. (female gland)Throat Dia. female gland D B Tube OD (female gland)G Groove width$ESIGN 'UIDE 'UIDE FOR $ESIGN 4ABLE

~~)NDUSTRIAL 3TATIC 3EAL 'LANDS Male Gland Female Gland 1/2 E 1/2 E B-1 Dia. (B-1 Min. = A Max. -2 L Max.) B Dia.~~

C Dia. D Dia.(e) A Dia. A-1 Dia.0° to 5° Break Corners (A-1 Max. = B Min. +2 L Max.)(Typ.)Approx. .005 RAD.1/2 E W W Pressure Pressure Pressure

~~.005 32 Typ.~~

W R W 63 F L Gland 63 Depth I.D.32

~~.003 Typ.~~

G G G1 G2 F Groove Section W-W No One Two Depth (Ref.)Gland Detail Parbak Parbak Parbak Finishes are RMS values. Ring Ring Rings Refer to Design Chart 4-2 (below) and Design Table 4-2 for dimensions metric conversion 323 = .83

)NDUSTRIAL / 2ING 3TATIC 3EAL 'LANDS ' 'ROOVE 7IDTH / 2ING 7 %A No /NE 4WO 2 -AX 3IZE #ROSS 3ECTION , 'LAND 3QUEEZE $IAMETRAL 0ARBAK 0ARBAK 0ARBAK 'ROOVE %CCENTRICITY!3" .OMINAL !CTUAL $EPTH !CTUAL #LEARANCE 2ING ' 2ING ' 2ING ' 2ADIUS B 004 .050 .015 22 .002 .093 .138 .205 .005 .070 +/-.003 through 1/16 to to to to to to to to .002 (1.78 mm) 050 .052 .023 32 .005 .098 .143 .210 .015 102 .081 .017 17 .002 .140 .171 .238 .005 .103 +/-.003 through 3/32 to to to to to to to to .002 (2.62 mm) 178 .083 .025 24 .005 .145 .176 .243 .015 201 .111 .022 16 .003 .187 .208 .275 .010 .139 +/-.004 through 1/8 to to to to to to to to .003 (3.53 mm) 284 .113 .032 23 .006 .192 .213 .280 .025 309 .170 .032 15 .003 .281 .311 .410 .020 .210 +/-.005 through 3/16 to to to to to to to to .004 (5.33 mm) 395 .173 .045 21 .006 .286 .316 .415 .035 425 .226 .040 15 .004 .375 .408 .538 .020 .275 +/-.006 through 1/4 to to to to to to to to .005 (6.99 mm) 475 .229 .055 20 .007 .380 .413 .543 .035 (a) Clearance (extrusion gap) must be held to a minimum consistent with design requirements for temperature range variation.

(b) Total indicator reading between groove and adjacent bearing surface.(c) Reduce maximum diametral clearance 50% when using silicone or "uorosilicone O-rings.D &OR EASE OF ASSEMBLY WHEN 0ARBAKS ARE USED GLAND DEPTH MAY BE INCREASED UP TO $ESIGN #HART &OR )NDUSTRIAL / 2ING 3TATIC 3EAL 'LANDS 0ARKER (ANNIlN #ORPORATION s / 2ING $IVISION 4-9 2360 Palumbo Drive, Lexington, KY 40509 0HONE s &AX www.parkerorings.com

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 13 of 66 REPRINTED FROM MACHINE DESIGN April 12, 1979 O-RINGS FOR LOW-PRESSURE SERVICE P.O. BOX 519 , RED BANK, NEW JERSEY 07701 (201) 747-9200

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 14 of 66 DANIEL L. HERTZ, JR.President Seals Eastern Inc. Red Bank, N.J.O-RINGS normally operate with about 15% squeeze to ensure a tight seal.But at system pressures below 400 psi, this amount of squeeze can cause high friction and excessively high actuating forces.Reducing the amount of squeeze lowers friction to acceptable levels; however, lower squeeze also means lower sealing pressure and greater potential for leakage. This problem is aggravated by the stress relaxation characteristics of the seal material.Thus, an O-ring that seals well initially may lose resilience with time and fail suddenly.Designing O-ring seals for low pressures, therefore, is not simply a matter of reducing the amount of squeeze: it involves a delicate balancing of material hardness, dimensional tolerances, stress relaxation, and friction characteristics.Material Hardness The initial phase of designing a low-pressure O-ring seal is the same as that for a conventional O-ring.Size and fluid compatibility requirements are evaluated and O-ring dimensions selected from a catalog. The catalogs usually list a recommended range of squeeze values, as shown in Table 1.Squeeze is defined as the ratio 2

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 15 of 66 Most O-rings operate with enough "squeeze" to provide a reliable seal under almost any conditions. But low-pressure systems generate less squeeze, increasing the potential for leakage. Only a careful balancing of O-ring material properties ensures leak-free operation at low pressures.with specifying squeeze at the low can be calculated from end of the range is that di- b = 2.4x mensional tolerances can reduce the amount of squeeze actually Then, the peak contact stress can placed on the O-ring. For instance, be found from tolerance on the 0.070-in. thick O-ring is +/-0.003 in. In the worst case (0.067-in. thickness), this tolerance can account for 8.6% of If f ' is greater than the system the squeeze allowance, leaving only pressure, the O-ring will seal the 6.3% to be supplied by the fit joint. If f ' is less than system between parts. In other words, an pressure, the ring will leak and a undersize O-ring has less material material with a higher Young's to compress and cannot be Modulus must be specified, squeezed as tightly against the thereby increasing compressive sealing surfaces. This problem can force and contact stress.be minimized by specifying O-rings with one-half the normal Seal Friction dimensional tolerances. Such seals In low-pressure systems, seal are available from most friction can raise the required manufacturers at a premium price. actuating pressure to many times The next step in the design that available in the system.procedure is to calculate the Therefore, seal friction must be compression force developed in minimized for the system to the O-ring. This force is directly operate properly. Generally, seal related to the sealing ability of the friction force should be maintained ring and is calculated from below 20 lb to keep actuating force within reasonable limits.The friction force for an O-ring seal can be estimated from 75 1340 To use this equation, Young's Modulus must be determined first.of seal deflection to seal thickness, This value depends on material where coefficient of friction, ,x/d. Generally, the seal is designed hardness, and typical values are can change from 0.001 to over 10, to operate at the high end of the listed in Table 2. For most depending on the operating squeeze range to ensure a tight applications, a Shore A hardness of conditions.seal. But at low system pressures, 70 is sufficient; therefore, the initial When more than one O-ring is squeeze must be specified at the calculation of F is based on this used in the system, the friction low end of the range.hardness. forces from all the seals must be The squeeze values listed in From the specified squeeze and combined to determine the total Table 1 are based on nominal seal seal thickness, contact area friction force. If the calculated thickness. One problem force is greater than

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 16 of 66 20 lb, a softer seal material force must be lowered by reduc- an unlubricated seal.should be used; this lowers ing the coefficient of friction. This increase with time is Young's Modulus and compres- This factor is a complex function caused by the atomic interaction sive force. However, the change of lubricant film thickness, time, between the O-ring and its to a softer seal material must be contact stress, sliding speed, and sealing surface, which causes the made with care because a lower surface finish. two surfaces to adhere tightly.compressive force also means a Tests have shown that the The adhesive force can be quite lower contact stress. Thus, the longer a lubricated seal sits idle, high and eventually squeezes change could lower peak contact the higher its static, or most of the lubricant from under stress below system pressure, breakaway, coefficient of friction. the contact area. On start-up, the resulting in a leaky seal. Eventually, the friction adhered O-ring peels away in If a softer material lowers coefficient reaches a maximum progressive waves that break contact stress too much, friction value almost as high as that for away and reform

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 17 of 66 on the moving surface, This action shears what little lubricant is present and traps it in the rubber folds.Seal adhesion can be minimized by optimizing surface finish and lubricant viscosity. Experience has shown that the optimum surface finish is 0.4 m. This finish leaves tiny pockets that collect lubricant, making it available at startup. Too smooth a finish leaves no pockets for the lubricant, while too rough a finish causes high wear.A reciprocating seal should be lubricated with high viscosity lubricants because they produce a strong hydrodynamic film. This film resists displacement by the adhesive forces when the seal is stationary. A rotary seal, on the other hand, can be lubricated with low-viscosity lubricants because rotary motion aids development of a hydrodynamic film.Thickness of the hydrodynamic film between asperities on the O-ring and sealing surface has been calculated as 6 x 10 in. Shear of this film is the prime cause of dynamic or running friction. In general, the dynamic coefficient of friction is a function of lubricant viscosity and sliding velocity. The coefficient generally starts high, decreases to a minimum value, then increases again.Thus, running friction can be minimized by optimizing viscosity and velocity.Several tests have been run to determine the effect of material-formula modifications on seal friction. The addition of materials such as graphite, molybdenum disulphide, and PTFE sometimes reduce friction, but the reduction is more likely a result of lowering Young's Modulus than a lubricating effect. Also, the incorporation in the elastomer of high-molecular-weight waxes and oils that migrate to the surface has proved unsuccessful in

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 18 of 66 Page intentionally blank

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 19 of 66 Table 3-Glass-Transition Temperatures for lowering friction. O-ring Materials Surface treatments have been Material Transition Temperature more successful. Halogenation with(°F) chlorine or bromine reduces friction Nitrile (NBR) 34% ACN -21 by lowering the surface free energy 38% ACN -13 (and, therefore, attraction force) and Fluoro Rubber (FKM) -4 by creating lubricant pockets. Silicone (VMQ) -85 Chloroprene (CR) -40 Fluorination, although far less Ethylene Propylene (EPDM) -85 common, has similar effects.Surface treatments of PTFE-resin with material composition, binder coatings and tumble temperature, and fluid reactions. steel. Therefore, at high temperatures, treatment in molybdenum disulphide Typical values range from 0.5% to insufficient groove volume can or graphite have been used along 10% per time decade. (The time from produce expansion forces that with silicone oil dips with limited 1 to 10 min is designated as one extrude the seal into the clearances.success. Also, polymerization of decade, as is the much longer time This problem can be minimized by monomers on the O-ring surface from 1 to 10 weeks.) increasing groove dimensions to with plasma techniques offers some The result of stress relaxation is provide sufficient room for improvement; however, the that peak compressive stress expansion.techniques are costly and slow. eventually drops below system Lowering operating temperature Finally, the grafting onto the seal pressure, and the seal leaks. Thus, results in a continuous decrease in the surface of high-molecular-weight oils stress relaxation effects must be physical volume of the seal.having reactive end groups shows factored into the determination of Eventually, the seal reaches its promise for the future. material hardness and compressive so-called glass-transition temperature, stress. where it seals only along two thin Stress Relaxation Stress relaxation rates are available lines. Further reduction of from O-ring manufacturers; temperature shrinks the seal even The useful sealing life of an however, the ratings may be for a more, resulting in leakage.O-ring depends on two viscoelastic temperature or fluid condition material properties: compression set, different from that required. If the The glass-transition temperature the residual deformation of a correct data are not available, the corresponds to 100% compression material after the load is removed; stress relaxation rate can be set. At this temperature, the seal and stress relaxation, the decrease in determined from a simple can shatter like glass if subjected to stress after a given time at a constant relaxometer test, such as that a shock or impact load. Values of strain. These properties reduce the described in ASTM D-1395, or with glass transition temperature for resiliency of the seal material and a Lucas relaxometer. O-ring materials are listed in must be taken into account when Once the stress relaxation rate is Table 3. To avoid low specifying material hardness. known, the time for peak contact temperature problems, O-rings When a seal is under constant stress to equal system pressure can be should operate at temperatures compression, the initial stress decays calculated easily. In general, if the 10° to 15°F higher than those at a rate proportional to the calculated time period produces a seal listed in the table.logarithm of time. The stress life lower than 20 x 106 cycles, then a relaxation rate varies harder seal material (higher Young's Modulus and higher compressive References Nomenclature stress) must be used. 1. C.J. Derham, "Elastomeric Sealing,"Engineering, May 1977.b = Seal contact area, in.2 Temperature Effects 2. P. B. Lindley, "Engineering Design with Di = Seal inside diam, in. Natural Rubber," Malaysian Rubber Dm = Seal mean diam, in. Producers' Research Association, London, D = Seal outside diam, in. The effects of operating 1974.d = Seal thickness, in. temperature are more pronounced 3. P. B. Lindley, "Compression E = Young's Modulus, psi for low-squeeze O-rings because the Characteristics of Laterally Unrestrained F = Compressive load, lb seal has a lower tolerance for change. Rubber O-Rings," Journal of the Institution, Ff = Friction force, lb of the Rubber Industry, July/August 1967.f = Peak contact stress, psi The volumetric expansion rate for x = Seal deflection, in. rubber is about 15 times higher than

~~4. A. D. Roberts, "Optical Rubber," Rubber~~
~~= Coefficient of friction that for Deuelopments, Vol. 29, No. 1, 1976.~~
~~5. A. D. Roberts, "Looking at Rubber Friction," Rubber Deuelopments, Vol. 29, No.~~

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 20 of 66 14th International Conference on Fluid Sealing, Firenze, Italy, 6-8 April 1994, Organised by BHRGroup Limited, Cranfield, Bedford, MK43 0AJ, UK STRESSES AND DEFORMATION OF COMPRESSED ELASTOMERIC O-RING SEALS by Itzhak Green and Capel English The George W. Woodruff School of Mechanical Engineering Georgia Institute of Technology Atlanta, GA 30332-0405 ABSTRACT The sealing capability of an elastomeric O-ring seal depends upon the contact stresses that develop between the O-ring and the surfaces with which it comes into contact. It has been suggested in the literature that leakage will occur when the pressure differential across the seal just exceeds the initial (or static) peak contact stress. The stresses that develop in compressed O-rings, in common cases of restrained and unrestrained geometries (grooved and ungrooved), are investigated using the finite element method. The analysis includes material hyperelasticity and axisymmetry. Contact stress profiles, and peak contact stresses are plotted versus squeeze, up to 32 percent. The contact width, which is the length of the O-ring that touches the retaining surfaces when viewed from the cross-section, is also determined. Expressions are derived empirically to predict the peak contact stress and the contact width. These expressions are also compared to those obtained by other researchers (who assumed plain strain conditions) and conclusions to their validity are drawn.NOMENCLATURE b = contact width S = compressive stress d = wire diameter x* = displacement D = nominal (mean) diameter x = radial coordinate Ddef = deformed mean diameter X = radial distance from O-ring center E = modulus of elasticity y = axial coordinate h = deformed O-ring thickness Y = axial distance from O-ring center l = groove width = normalized squeeze (i.e., fractional q = chord diameter compression)Q = normalized chord diameter, q/d ij = equivalent normalized squeeze 83

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 21 of 66 84 INTRODUCTION Elastomeric O-ring seals have a broad range of service conditions that make the O-ring ideal for static and dynamic sealing functions. Its ability to seal on relatively rough surface finishes offers one of the economical solutions to sealing problems. Elastomeric O-rings are capable of undergoing large deformations under compression. Hence, grooves are often used to restrict this deformation, resulting in improved sealing capabilities and prevention of creep and extrusion. The complex geometry confederated with the deformation of restrained O-rings and nonlinear material hyperelasticity render analytical solutions infeasible. This complicated geometry and experimental inconvenience make experimental data hard to obtain. It is here where the utility of the finite element method becomes prominent. By performing a FEM analysis, comparison of the results can be made to cases where experimental data is procurable. Then, conclusions can be drawn as to the validity of FEM solutions of geometries where experimental data cannot be easily obtained.The stiffness relationships associated with the compression of elastomeric torroidal O-ring seals have recently been studied by Green and English (1992) for the cases shown in Figure 1. That work provided empirical expressions for the prediction of compression forces and stiffnesses at squeeze levels up to 32 percent. Sealing capabilities, however, depend upon the stress related parameters at the interface.It was as early as Lindly's work (1967), Figure 1 (A) Unrestrained Radial Loading. (B) who proposed that leakage onset occurs Unrestrained Axial Loading. (C) Restrained when the pressure differential across the Radial Loading. (D) Restrained Axial loading.seal, P, barely exceeds the initial (or static) peak contact stress, Smax (i.e., P Smax)). It should be noted that any increase in the contact stress, caused by the pressure loading, is ignored using this theory. Simplified expressions relating contact width to peak contact stress have been developed in order to predict Smax. Assuming unrestrained loading and plain strain Lindley (1967) obtained the contact width, b, normalized with respect to the wire diameter, d, [see Figures 2(a) and 3]Figure 2(a) Unrestrained geometry Figure 2(b) Section of a restrained O-ring

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 22 of 66 85

~~(1)~~

F %and the peak contact stress, Smax, normalized with respect to the modulus of elasticity, E,

5OCZ

~~(2) ' %~~

Here =x*/d, is the normalized squeeze, i.e., the compressive displacement, x*, divided by the wire diameter, d. The first term in the equations was obtained using Hertzian theory, and the second term was added to correct for empirical data at high squeeze levels.Wendt (1971) examined stress distributions in O-rings and X-rings, with emphasis on groove design. The most significant result of his work includes an expression for contact width of an unrestrained axially loaded O-ring. Molari (1973), who examined the stress and contact related Figure 3 Stress profile definition for parameters of O-rings using photoelastic unrestrained axial loading.techniques, lent credence to the findings of Wendt and was one of the firsts to examine the problem of restrained O-ring seals. Molari's work, however, considered one lateral wall only. Dragoni and Strozzi (1988) also used photoelasticity, but investigated an O-ring restrained between two lateral walls as defined in Figure 2(b). These researchers assumed that plane strain conditions were prevailing and thus did not address the condition of axisymmetric loading.George, Strozzi, and Rich (1987) supported Lindley's results using a finite element code developed especially for the task. Experimental data taken was compared to the results obtained by numerical solution. Later Dragoni and Strozzi (1988) examined the case of laterally restrained O-ring seals in a groove using a modification of the FEM code. The results were also limited to plain strain conditions. Using Lindley's model of Hertzian contact stress, they offered an approximate analytical method for "moderately" compressed O-rings up to about 15 percent squeeze. A stress related parameter was given in terms of a normalized deformed chord diameter, Q = q/d (see Figure 3). By fitting a curve to experimental results (Strozzi, 1986), they characterized Q as a function of 3

~~H (3) where the right-hand side emphasizes the functional form of the equation, as needed for later derivations. Using only the first term of Eq. (2) the peak contact stress was given as~~

5OCZ (4)

~~' %~~

In a compromise between accuracy and simplicity they prefer Wendt's (1971) description of the contact width

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 23 of 66 86

D (5)F At this point Dragoni and Strozzi (1988) developed an equivalent normalized squeeze, ij, which can be used in modeling the characteristics of restrained O-rings. The notation ij is used to denote the effects of the particular groove wall defined perpendicular to either the i or j direction.For example, for restrained radial loading the equivalent squeeze in the x direction, xy, denotes the squeeze associated in part to the squeeze directly applied in the x direction and in part to the constraint of the walls which are perpendicular to the y direction. Alternately, for restrained axial loading, yx is the equivalent normalized squeeze on the groove walls perpendicular to the top/bottom compressive surfaces. By definition yx is estimated as a ratio. The numerator is the difference between two terms: (i) a virtual deformed chord diameter along the y-axis caused by the compression xy [and is calculated by substituting xy into Eq. (3)]; (ii) the deformed O-ring thickness, h [shown in Figure 2(b)]. The denominator is the undeformed wire diameter, d. Hence, F#H Z[ J J[Z H Z[ (6)F F where f is the functional given in Eq. (3). Applying similar reasoning in the perpendicular x-direction, and using the groove width, l (as the O-ring thickness), gives the equivalent squeeze F#H [Z N N Z[ H [Z (7)F F These relationships provide estimates for any groove dimensions (allowing the possibility of a gap between the undeformed O-ring and the lateral walls, i.e., l > d). Next, we define the particular case (subscripted here with the letter t) where the groove lateral walls are tangent to the undeformed O-ring, i.e., l = d as shown in Figure 2(b). Combination of Eqs. (6) and (7) yields JV[Z H H V[Z (8)F where the functional form of Eq. (3) is used repeatedly. Eq. (8) can be solved iteratively for tyx.Then txy is calculated by Eq. (7), and by substitution into Eqs. (4) and (5) the normalized peak contact stress and the normalized contact width can be determined in the respective directions.Since in all the aforementioned work plain strain conditions prevailed, it implies that no distinction exists between axial and radial loading. This was found invalid in some important loading conditions for the compression force and stiffness (Green and English, 1992). The loading cases in Figure 1 are investigated here to determine contact stresses and contact widths under axisymmetric conditions. These include a highly frictional ("unlubricated") contact where surface sliding is prevented in an unrestrained axial loading; and frictionless ("perfectly lubricated") contacts where forceless surface sliding exist in axial, radial, restrained, and unrestrained loadings. The commercial code ANSYS and the nonlinear techniques, described in Green and English (1992) and in greater detail in English (1989), are utilized. Reduced integration is exclusively applied as it was found to give most accurate results. These are best represented in normalized forms, proven indifferently to the aspect ratio, d/D. Convergence is discussed in whole in the last two references.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 24 of 66 87 STRESS PARAMETERS RESULTS Vernacular for this discussion includes "primary wall" and "lateral wall." The primary wall or walls are the surfaces which move together to force the compression of the O-ring. The lateral walls are the sides of the restraining groove. Initially all walls are tangent to the undeformed geometry of the O-ring. For the axial case the top and bottom walls are the primary walls, and for the radial case the inside and outside walls are the primary walls.Contact stress profiles are plotted as the normalized nodal stress component, Sy/E or Sx/E, versus the normalized x or y coordinates (X/d or Y/d), respectively. For example, Figure 3 shows the X-coordinate, defined relative to the deformed nominal radius, Ddef/2. X is the horizontal distance from the center of the O-ring cross-section to a node along the perimeter. The normalized stress of interest here is the y-component of the nodal stress, Sy/E. In the plots Eo is shown instead to designate neo-Hookean material representation, justified for use by Green and English (1992).The first contact pressure profile, shown in Figure 4, was produced from a quadrilateral element mesh, using reduced integration (Green and English, 1992). Six profiles were chosen Figure 4 Primary wall contact stress profile Figure 5 Primary wall contact stress profile for for unrestrained - perfectly lubricated unrestrained - unlubricated axially loaded O-axially loaded O-ring. ring (the only fixed case) from ten load steps of 3.2 percent squeeze each, and are represented by the symbols given in the legend. Two other features can be extracted from the figure: (i) The normalized deformed chord diameter, Q=q/d, is the farthest distance between two opposite points on each profile, and (ii) the normalized contact width, b/d, is the distance between two points on each profile where the curve intersects the zero stress line. Notice that Q and b/d are different at each load step. Finally, there is a condition of symmetry about the x-z plane, of the global coordinate system, such that both the top and bottom profiles are identical and, therefore, only one is shown. While the profiles appear symmetric about X/d=0, close examination of the results reveals otherwise. This is due to the imposition of axisymmetric conditions upon the solution (rather than plain strain conditions).Figure 5 shows the contact stress profile for the unlubricated case of unrestrained axial loading (where a coefficient of friction 0.9 was used). The asymmetry is much more pronounced in

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 25 of 66 this case. It should also be noted that the peak value of Sy is roughly 85 percent higher than that for the lubricated case. This increase in peak contact stress can be explained by the fact that the deformed nominal diameter, Ddef, does not expand as the load increases. Actually Ddef decreases as the loading is applied, although, only a small amount. A comparison of Figures 4 and 5, shows that friction has a dominating role in stress profile development.Turning to radial compression, Figure 6 shows how the contact stress profiles are defined.Sx is the stress component of interest here, and Y is the vertical distance from the center of the Figure 6 Stress profile definition for Figure 7 Contact stress profile for unrestrained radial loading. unrestrained radial loading. Half profiles for the inside and outside primary walls are shown.O-ring cross section to a node on the perimeter. q is the deformed chord diameter parallel to the compressive surfaces. Because the model is not symmetric about the y-z plane the stresses on both walls, the inside primary wall and the outside primary wall, must be examined.The stress profiles for unrestrained radial loading are shown in Figure 7. Here we notice that the outside primary wall profile is larger than the inside primary wall profile. This is due to nominal diameter contraction during axisymmetric loading. Figure 7 exemplifies again the necessity of using an axisymmetric model to represent the O-ring. While the difference between the inside and outside walls is minor for this particular restraining configuration it could be much more significant for a different type of loading.Next up for attention are the restrained cases. Figure 8 gives the profile definition for restrained axial loading. Note that the profiles are symmetric about the x-z plane; however, they Figure 8 Profile definition for restrained Figure 9 Profile definition for restrained axial loading. radial loading.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 26 of 66 89 are not symmetric about the y-z plane. The restrained radial profile definition can be seen in Figure

9. This profile is similar to the previous profile in that the symmetry planes are the same. However, the primary and lateral walls are different. Note the definitions for the contact widths b and c on the primary and lateral walls, respectively.

Figures 10 and 11 show the profiles for the primary and lateral walls, respectively, for the restrained axially loaded O-ring. The half profiles for the inside and outside lateral walls show Figure 10 Contact stress profile for Figure 11 Contact stress profile for inside and primary wall of restrained axial loading. outside lateral walls for restrained axial loading.that there is a difference between lateral wall profiles where axisymmetric loading is concerned.Here the peak value of S/Eo is 17 percent larger for the primary wall than that for the lateral wall.Figures 10 and 11 show an interesting formation of significant compressive normal stresses at the O-ring surface that is in contact with the respective retaining walls. Molari (1973), using bidimensional photoelastic techniques, obtained similar profiles; however, the surface stresses did not show up in his work because they were masked by the physical boundary of the test apparatus.Figure 12 Contact stress profile for inside Figure 13 Contact stress profile for lateral and outside primary walls for restrained wall of restrained radial loading.radial loading.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 27 of 66 90 Next we consider the case of restrained radial loading. Contact stress profiles for the primary and lateral walls are given in Figures 12 and 13, respectively. In Figure 12 a difference exists in the half profiles for the inside (left) and outside (right) primary walls.Peak contact stress values for the primary walls are about 34 percent greater than those for the lateral wall. Also here there are surface stresses at the retaining walls.The final case under investigation is that for plane strain loading. Figure 14 shows the stress profile for unrestrained plane strain loading. Plane strain loading profiles for the restrained case are given in Figures 15 and 16.In this case there is symmetry about both the x-z and y-z planes. Hence, there is no need for a Figure 14 Contact stress profile for half profile plot of inside or outside walls. It unrestrained plane strain loading primary wall.can be seen by comparing the stress profiles from plain strain loading to all previous cases that there is a significant difference in the profiles, especially in radial loading and peak contact stresses. It is, therefore, concluded that plain strain conditions do not commonly describe O-ring compression.Figure 15 Contact stress profile for Figure 16 Contact stress profile for plane restrained plane strain loading primary wall. strain loading lateral walls.PEAK CONTACT STRESS Peak contact stress is of interest in order to estimate the ability of the O-ring to form a seal (Lindley, 1967). Figure 17 contains the compilation of peak contact stresses for unrestrained loadings, i.e., axial, radial, and plane strain. It can be seen that the primary wall peak contact stress response for radial loading is the greatest. This is followed by the stress response of the unlubricated axial loading, which is significantly greater than its lubricated correlative. Furthermore, we see that both Lindley's predictions underestimate the peak contact stress throughout the loading range with the exception of the plane strain case. The latter case agrees relatively well with the prediction Lindley derived from Hertzian theory (Eq. (2), but without the second correction term). Note that

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 28 of 66 91 the empirically added correction term causes an overestimation of the peak contact stress for squeezes above 24 percent.Lindley gives no prediction for the peak contact stress of a restrained O-ring.The comparison to Dragoni and Strozzi (1988) prediction can be seen in Figures 18 and 19. Their prediction agrees relatively well with the axial and plane strain results, but, underestimates the peak contact stress for radial loading. Where the lateral wall is concerned Strozzi's prediction underestimates the peak contact stress response for all cases. It is interesting that the lateral wall response for radial loading is less than the responses for both axial and plane strain cases. It seems from this Figure 17 Peak contact stress results for comparison, that Strozzi's model is relatively unrestrained loading.accurate for predicting the peak contact stress for the primary wall of restrained axial and restrained plane strain cases, but it breaks down in radial loading and where the lateral wall is concerned.Figure 18 Peak contact stress results Figure 19 Peak contact stress for restrained for restrained loading primary wall. loading lateral wall.The lack of agreement to analytical work prompts the determination of the peak contact stress from the numerical data. This is accomplished by fitting a polynomial to the numerical results.Second and third order polynomials are proposed as follows:5OCZ C D (9) 5OCZ E F

(10)

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 29 of 66 92 where the coefficients are given in Table 1. While Eq. (10) produces an excellent fit, Eq. (9) may be used for simplicity with satisfactory results.Table I Least squares coefficients for the calculation of the peak contact stress Loading Case a b c d e Unrestrained lubricated axial primary wall 2.0572 -3.1417 2.6296 -8.8589 12.8391 Unrestrained unlubricated axial primary wall 2.0090 -0.2211 2.8383 -8.5051 18.6031 Unrestrained lubricated radial primary wall 2.4891 -1.5967 3.4591 -11.2857 21.7583 Unrestrained lubricated plane strain primary wall 2.2340 -2.8961 3.0373 -10.9192 18.0171 Restrained axial primary wall 1.9715 3.1502 3.8295 -23.0013 82.6963 Restrained axial lateral wall 1.0497 6.4631 2.6584 -16.1793 71.5999 Restrained radial primary wall 2.1587 6.7729 4.9363 -32.3232 123.630 Restrained radial lateral wall 0.5844 8.7930 2.3003 -15.3593 76.3744 Restrained plane strain primary wall 1.9933 3.2711 4.0499 -25.6765 91.5384 Restrained plane strain lateral wall 0.9400 7.5182 2.6698 -16.8290 76.9908 CONTACT WIDTH RESULTS In Figure 3 the contact width, b, is defined as the length of the circumference of the O-ring, from a cross-sectional view, that makes contact with the compressing surface.This information is useful in calculating the total load required to compress the O-ring and in determining the contact stress profile when using Hertzian theory. Both Lindley (1967) and Wendt (1971) propose expressions which approximate the contact width as a function of compression for unrestrained loading, and Dragoni and Strozzi (1988) propose corresponding expressions for restrained loading. As illuminated in the introduction these were developed assuming plain strain conditions. This section compares results obtained in this research to those predicted by the other researchers. Figure 20 Normalized contact width as a The error associated with the numerical function of compression for unrestrained results is significant when considering the loading. Included are Lindley and Wendt's technique used to obtain the contact width. In prediction as well as the results from nodal Figure 20 there are several data points with displacements.approximately identical values of b/d. This comes from the discrete points used to obtain the contact width. As the loading is applied, new nodes may or may not come into contact with the compressing surfaces. Several data points with the same value of b/d imply that no new nodes have come into contact during that portion of the loading sequence. Physically the contact width response is a continuous phenomenon. By discritizing the mesh we turn this response into a discrete phenomenon. Consequently the first data point, of those which have the same magnitude of b/d, is

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 30 of 66 93 the most accurate. Therefore, the contact width shown is an underestimation of the actual contact width.Contact width data, for the cases of unrestrained loading can be seen in Figure 20. The numerical results from the unlubricated axial and radial cases agree well with Wendt's prediction, Eq. (5), up to roughly 24 percent compression. Lindley's prediction, Eq. (1), underestimates the numerical contact width throughout the load range. For the lubricated axial load case Wendt's expression begins to overestimate the contact width at approximately 10 percent compression while Lindley's expression underestimates, at low compression, and over estimates it at higher compressions. From this we may conclude that Wendt's prediction is good for small compressions and that Lindley's prediction should generally not be used.Now we turn to restrained loading while probing Strozzi's approach as outlined from Eq. (3) through Eq. (8). To do so a second order polynomial [similar to Eq. (3)] must first be fitted to the data obtained from the FE analysis of the three unrestrained loading results (lubricated and unlubricated axial loading, and radial loading). By extracting nodal displacements from the output it is possible to obtain the normalized deformed chord diameter. For the unlubricated-unrestrained axial loading case the fitted polynomial is 3 (11)For the lubricated-unrestrained axial loading case the polynomial obtained is 3 (12) and for the unrestrained radial loading case the polynomial obtained is 3 (13)The contact width results for restrained loading can be examined in Figures 21 and 22 for the primary wall and lateral walls, respectively. Looking at Figure 21 we see that the predictions Figure 21 Contact width as a function Figure 22 Contact width as a function of of compression for the primary wall compression for the lateral wall of a restrained of a restrained O-ring. O-ring.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 31 of 66 94 in the solid lines, using Eqs. (12) and (13) as well as Strozzi's Eq. (3), produce overestimates compared to the numerical nodal contact width at the primary wall for compressions below 19 percent. At the extreme compressions the normalized chord diameter technique underestimates the contact width of the primary wall of a restrained radially loaded O-ring. However, as mentioned above, these numerical data points underestimate the actual contact width. Therefore, the prediction may be considered to give a fairly accurate prediction of contact width.In Figure 22 we see that Strozzi's technique is a closer approximation to the lateral wall contact width of a restrained axially loaded O-ring. But, the figure also shows that the technique is a gross underestimate of the lateral wall contact width for a restrained radially loaded O-ring.Without more experimental results it is hard to make a firm statement as to the accuracy of either Strozzi's approach or the current numerical approach. It can generally be said, however, that the numerical results underestimate the actual contact width and, therefore, the method Strozzi suggests is valid, with exception to the radially loaded case at the lateral wall.CONCLUSIONS Acquisition of the stress parameters requires more effort in the postprocessing phase of a finite element analysis, but compared to the extensive testing equipment required for experimental stress data acquisition, this methodology is far less expensive in terms of resources and time.Another feature that FEA has to offer is the ability to examine surface stress data that is otherwise hidden by the boundary of experimental apparatus.The most profound finding of this work is the identification of a significant difference between the peak contact stress response of plane strain models and axisymmetric models of hyperelastic O-rings (see Figs. 17 through 19, and Table 1). This is particularly true for the case of unrestrained loading . The plane strain results obtained agreed well with those predicted by Wendt.However, Wendt's prediction of peak contact stress response greatly underestimated the response generated by axisymmetric loading. Results obtained from axisymmetric modeling of the lubricated-axially loaded O-ring also indicate that the plane strain assumption is not valid for prediction of the peak contact stress response for this particular case.Contact width examinations performed in this work yield the most inconclusive results out of all the topics investigated. This is primarily because the mesh is finite at the perimeter and only discrete information about the contact width is available. Clearly better results can be obtained with a much refined mesh at the expense of computer time. Given the results for unlubricated-unrestrained axial loading reasonable agreement exists with Wendt's prediction for compressions up to 15 percent. Lindley's prediction underestimates the contact width response for all cases except lubricated-unrestrained axial loading.Looking at the cases of restrained loading, only Strozzi offered an analytical technique of predicting contact width and stress. To conform with that technique the numerical results were fitted to produce expressions for the normalized chord diameter in Eqs. (11) - (13). However, the use of these equations, in the procedure outlined from Eq. (3) through Eq. (8), overestimates the contact width compared to numerical data obtained from the deformed nodal coordinates. Strozzi's prediction underestimates the lateral wall stress response for the axisymmetric axial and plane strain cases, yet it gives relatively good agreement to the axisymmetric radial case for compressions below 10 percent. For compressions above 10 percent Strozzi's prediction again underestimates the peak contact stress response.Due to the lack of consistent agreement between the results obtained here and previous analytical work an alternate empirical procedure is proposed. The peak contact stresses can be determined using Eqs. (9) or (10) for the ten loading conditions listed in Table 1. According to Lindley (1967) these equations provide estimates of the maximum pressure an O-ring can seal.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 32 of 66 95 ACKNOWLEDGMENT The authors gratefully acknowledge the support given to this work by the National Science Foundation REU Program under grant number MSM-8619190.REFERENCES Dragoni, E., and Strozzi, A., 1988, "Analysis of an Unpressurized Laterally Restrained, Elastomeric O-ring," Trans. ASME, Journal of Tribology. Vol. 110, No. 2, pp. 193-199.English, C., 1989, "Stiffness Determination of Elastomeric O-Rings Using the Finite Element Method," M.S. Thesis, Georgia Institute of Technology.George, A.F., Strozzi, A., and Rich, J.I., 1987, "Stress Fields in Compressed Unconstrained Elastomeric O-ring seals and a Comparison with Computer Predictions with Experimental Results,"Tribology International Vol. 20, pp.237-247.Green, I., and English, C., 1992, "Analysis of Elastomeric O-ring Seals in Compression Using the Finite Element Method," STLE Trib. Trans., Vol. 35, No. 1, pp. 83-88.Lindley, P.B., 1967, "Compression Characteristics of Laterally Unrestrained Rubber O-ring," J. IRI, Vol. 1, pp. 202-213.Molari, P.G., 1973, "Stresses in O-ring Gaskets," 6th Int. Conf. on Fluid Sealing BHRA, pp. B2/15-31.Strozzi, A., 1986, "Experimental Stress-Strain Field in Elastomeric O-ring seals," Experimental Stress Analysis (H. Wieringa ed.), Martinus Nijhoff Publ., pp. 613-622.Wendt, G., 1971, "Investigation of Rubber O-ring and X-rings, 1. Stress Distributions, Service and Groove Design," BHRA Fluid Engineering, Internal Report T1115.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 33 of 66 O-ring Seal Design Best Practices 12-15-12 Rev. 1 1 of 7

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 34 of 66 Table of Contents 1.0 O-RING SEALS - THEORY AND DESIGN PRACTICES ......................................................................... 3 2.0 ANALYSIS OF O-RING SEAL DESIGNS ................................................................................................... 6 3.0 APPENDIX ................................................................................................................................................... 7 2 of 7

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 35 of 66 1.0 O-RING SEALS - THEORY AND DESIGN PRACTICES Theory:An o-ring seal consists of an o-ring and a properly designed gland which applies a predictable deformation to the o-ring. The gland is basically a groove dimensioned to a certain height H and width W (Figure 1) to allow a fixed compression of the o-ring when the gland flanges make metal to metal contact. It is also oversized volumetrically such to allow accommodation of the o-ring as it flows under compression. Unlike gaskets which seal just by the resiliency of the material under mechanical compression of the joint, an o-ring can provide a seal both through the resiliency of the pre-compressed material and the pressure activation of the seal. The pre-compression of the o-ring applies a calculated mechanical contact stress or pressure at the o-ring contacting surfaces in the gland. As the o-ring seal is pressurized or activated the pressure on the o-ring further increases the contact stress on the o-ring contacting surfaces of the gland as the o-ring moves or flows toward the low pressure side. This means the pressure of the contained fluid transfers through the essentially incompressible o-ring material, and the contact stress rises with increasing pressure.As long as the pressure of the fluid does not exceed the contact stress of the o-ring, leakage should not occur.Figure 1 - O-Ring Groove Dimensions At zero gauge pressure, only the pre-compressed resiliency of the o-ring provides the seal (see Figure 2). If the system pressure is in the range of 0 - 100 or even to 400 pounds per square inch (psi) it can be considered as low pressure (in this paper 400 psi or less is considered low pressure), and the seal is maintained predominantly by the pre-compression or squeeze on the o-ring and its resulting contact stress. As the pressure increases, the o-ring is forced to the low pressure side of the gland. This provides an additional increase in contact stress as the o-ring deforms to a D shape (see Figure 3) and the contact area of sealing under pressure increases to almost twice the original zero-pressure area. For this reason, an o-ring can easily seal a high pressure as long as it does not mechanically fail.3 of 7

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 36 of 66 Figure 2 - O-ring un-pressurized Figure 3 - O-ring with high pressure Best Design Practices:

Figure 4 shows a poor surface finish and affect the tool mark direction. Such a finish can cause leaks.Figure 4 - Surface Finish Can Cause Leaks

~~b. As for o-ring compression or squeeze it is a result of three factors: the force to compress the o-ring, durometer, and cross section thickness. A 15-20%~~

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 37 of 66

The larger the cross-section, the more effective the sealing and longer the life of the seal. However, with a dynamic application in which friction is a factor, a compromise will be required.

5 of 7

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 38 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 39 of 66 sealing system, a margin above that is prudent, so it is suggested to follow the best practices in Section 1.0 in order to achieve a successful seal.Graph 1 - Contact Stress vs. Compression (Lindly) 1100 1000 900 Durometer - Shore A 800 60 70 80 Contact Stress- PSI 700 600 500 400 Low Pressure Zone 300 200 100 05 10 20 30 40

7 of 7

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 40 of 66 (TVA response to KEI data request 7-16-2020)Lift check valves Seat Component Id contact Group Drawing No. diameter Spring OD/Spring preload 32-2 1-CKV-32-293 32-2 1-CKV-32-303 TVDD9911X01-2 2 1/2" Design not available 32-2 1-CKV-32-313 43-1 1-CKV-43-834 43-1 1-CKV-43-841 N89-180-1-MD-82013603 43-1 1-CKV-43-883 43-1 1-CKV-43-884 Dwg. states 2-4 psid 61-1 1-CKV-61-533 Dwg. states a 5 psi spring 61-1 1-CKV-61-680 installed on earlier valves; latest 1-CKV-61-692 spring is a 10 psi cracking 61-1 W9825144 pressure - use conservative 1-CKV-61-745 value since either valve could be 61-1 installed.1-CKV-62-639 7500001295 TVD-D-9556 Dwg. TVD-D-9556 states 4 psi 62-4 1 3/8" cracking Spring part no. CatId is 1-CKV-63-868 TVW1-30608GS-(2) CWT472P, and PEG is reviewing 63-1 1" doc for characteristics.67-1 1-CKV-67-575A 67-1 1-CKV-67-575B W9825144-MD-67-1 1-CKV-67-575C 00403336 67-1 1-CKV-67-575D 1" same as group 61-1 67-3 1-CKV-67-585A Dwg. is TVW-D-30504-2(U1) and 67-3 1-CKV-67-585B 7257697/TVW-D-30504-2 for U2 VTD-K085-0090 67-3 1-CKV-67-585C Same spring as group 63-1 w/67-3 1-CKV-67-585D PEG searching.dwg. has 0.5 psi (max.) cracking 1-CKV-68-849 30508GLS-(2) pressure - note -NO SPRING -68-1 1 3/8" SOFT SEAT

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 41 of 66 From: Neal Estep To: Sandhya Shankar Cc: Mital Mistry

~~Subject:~~

FW: valve info request for LLRT design change Date: Monday, August 3, 2020 6:22:17 AM

Sandhya, Attached is some additional information for you piston/lift check valves.
Regards, Neal Neal Estep Kalsi Engineering, Inc. Main Office:

4410 Mint Hill Village Lane, Suite 201 745 Park Two Drive Mint Hill, NC 28227 Sugar Land, TX 77478 704-831-8950 - Charlotte Office/Direct 281-240-6500 - Main Switchboard 704-942-6773 - Mobile From: Cetta, William Frederick II <wfcetta5@tva.gov>Sent: Monday, August 3, 2020 7:19 AM To: Neal Estep <nestep@kalsi.com>Cc: Ortiz, Jose J <jjortiz@tva.gov>; Gowin, Mark Allen <magowin@tva.gov>

~~Subject:~~

FW: valve info request for LLRT design change EXTERNAL EMAIL: Do not open attachments or click on links unless you know the content is safe.See below for priority valve Flowserve information.

~~Thanks, William Cetta Design Engineer Mechanical Design Teleworking from home 865-335-1974 (Cell -primary contact) 423-365-1153 (Office)~~

TVA Watts Bar Nuclear wfcetta5@tva.gov From: Smith, John <JohSmith@Flowserve.com>Sent: Friday, July 31, 2020 2:37 PM To: Cetta, William Frederick II <wfcetta5@tva.gov>

~~Subject:~~

RE: valve info request for LLRT design change This is an EXTERNAL EMAIL from outside TVA. THINK BEFORE you CLICK links or OPEN attachments. If suspicious, please click the Report Phishing button located on the Outlook Toolbar at the top of your screen.

Bill, Group 63-1 Dwg. TVW1-30608GS-(2) for cracking pressure and seat contact diameter. Seat contact diameter was obtained from warehouse measurements. Cracking Pressure 2psi. Since this is a Soft seated valve it almost impossible to determine the seat contact: approx. 0.75 dia Group 70-1 Dwg. 13-103681-001 for seat contact diameter. Since this is a Soft seated valve it almost impossible to determine the seat contact: Approx. 3.03 dia
Best, John C. Smith Sales Engineer Flowserve Corporation johsmith@flowserve.com 919-271-6311 From: Cetta, William Frederick II <wfcetta5@tva.gov>

Sent: Tuesday, July 28, 2020 1:02 PM To: Smith, John <JohSmith@Flowserve.com>Cc: Ortiz, Jose J <jjortiz@tva.gov>; Gowin, Mark Allen <magowin@tva.gov>

~~Subject:~~

[External] RE: valve info request for LLRT design change CAUTION: This email originated from outside of Flowserve. Do not click links or open attachments unless you can confirm the sender and know the content is safe.Hello John, I was wondering on the status of the 2 priority items.Attached is previous information I sent related to the 2 items.Typically the cracking pressure is on a drawing, but the attached drawing TVW1-30608GS-(2) is missing this information. I am sending a PO of a spring for this valve we bought to assist.Here are the high priority valves for Flowserve data:Group 63-1 Dwg. TVW1-30608GS-(2) for cracking pressure and seat contact diameter. Seat contact diameter was obtained from warehouse measurements.Group 70-1 Dwg. 13-103681-001 for seat contact diameter.

~~Thanks, William Cetta Design Engineer Mechanical Design Teleworking from home 865-335-1974 (Cell -primary contact) 423-365-1153 (Office)~~

TVA Watts Bar Nuclear wfcetta5@tva.gov From: Cetta, William Frederick II

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 42 of 66 From: Cetta, William Frederick II To: Neal Estep; Ortiz, Jose J; Driskell, Charles Edmond Cc: Sandhya Shankar; Mital Mistry; Gowin, Mark Allen

~~Subject:~~

RE: Hard Seat of Soft Seat TVDD9911X01-2 Date: Sunday, August 16, 2020 1:42:30 PM Attachments: Appendix A - Unit 1 Scope and Drawing List wfc cmnts.pdf sys 32 chk vlv soft seat 55852.pdf TVD-D-991X01-(2) mk no 47W555-25A.pdf Appendix B - Unit 2 Scope and Drawing List wfc cmnts.pdf WBN-TVD-D-9911-(2)-1-MD-86630_U2 w soft seat.pdf EXTERNAL EMAIL: Do not open attachments or click on links unless you know the content is safe.Based on detail review, valves were confirmed to be soft seat as per previous direction.We will add the change paper to the App A drawing list input information to supplement the U1 dwg. w/ no soft seat noted as a comment to Jose and Charlie.U2 needs no change paper - We are changing the U2 drawing list App B to reflect the u2 drawing which has the soft seat on the drawing.Also, I noticed for Unit 1 (Appendix A) specified drawing no. for group 32-2 applicable to 1-CKV 293 and -313 was not correct - it it should be TVDD9911-(2) and needs change paper since this dwg.version (U1) does not reflect the soft seat. Valve is same configuration as the u2 drawings - i.e. no new configuration. All are soft seat assemblies.Jose/Charlie - The U1 group 32-2 check valves have 2 different drawings based on 2 different mark nos. per piping dwgs. and Maximo. Please correct the Appendix A prior to signing. (Reference old DCN no. 38814). Also, the App B group 32-2 drawing has a U2 w/ soft seat for these 3 valves (see attached - revise App B accordingly.Thanks, William Cetta Design Engineer Mechanical Design Teleworking from home 865-335-1974 (Cell -primary contact) 423-365-1153 (Office)TVA Watts Bar Nuclear wfcetta5@tva.gov From: Cetta, William Frederick II Sent: Tuesday, August 04, 2020 2:17 PM To: Neal Estep <nestep@kalsi.com>Cc: Sandhya Shankar <SShankar@kalsi.com>; Mital Mistry <mcmistry@kalsi.com>; Gowin, Mark Allen <magowin@tva.gov>; Ortiz, Jose J <jjortiz@tva.gov>

~~Subject:~~

RE: Hard Seat of Soft Seat TVDD9911X01-2 I am confident it is a soft seat - This is a case of change paper - I will send more supporting documentation.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 43 of 66 Thanks, William Cetta Design Engineer Mechanical Design Teleworking from home 865-335-1974 (Cell -primary contact) 423-365-1153 (Office)TVA Watts Bar Nuclear wfcetta5@tva.gov From: Neal Estep <nestep@kalsi.com>Sent: Tuesday, August 04, 2020 2:13 PM To: Cetta, William Frederick II <wfcetta5@tva.gov>Cc: Sandhya Shankar <SShankar@kalsi.com>; Mital Mistry <mcmistry@kalsi.com>; Gowin, Mark Allen <magowin@tva.gov>

~~Subject:~~

Hard Seat of Soft Seat TVDD9911X01-2 This is an EXTERNAL EMAIL from outside TVA. THINK BEFORE you CLICK links or OPEN attachments. If suspicious, please click the Report Phishing button located on the Outlook Toolbar at the top of your screen.Bill, I was wondering if you could verify that the Group 32-2 Kerotest check valves are SOFT seat. If so, could you please a suitable reference. The drawing TVD-D-9911-(2) shows a soft seat option in View B, but drawing TVDD9911X01-2 only shows a hard seat.The original spreadsheet we received from Jose for Unit 1 indicated they were soft seated, but the drawings are unclear.Unit 1 list:TVDD9911X01-1-CKV-32-293 CONTROL AIR CNTMT CHECK 32-2 2 TVDD9911X01-1-CKV-32-303 ESSENT CNTL AIR CNTMT CHECK 32-2 2 TVDD9911X01-1-CKV-32-313 ESSENT CNTL AIR CNTMT CHECK 32-2 2 Here is the Unit 2 list:TVDD9911X01-2-CKV-32-323 ESSENT CNTL AIR CNTMT CHECK 32-2 2 TVDD9911X01-2-CKV-32-333 ESSENT CNTL AIR CNTMT CHECK 32-2 2 TVDD9911X01-2-CKV-32-343 CONTROL AIR CNTMT CHECK 32-2 2 Thanks,

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 44 of 66 Neal Neal Estep Kalsi Engineering, Inc. Main Office:4410 Mint Hill Village Lane, Suite 201 745 Park Two Drive Mint Hill, NC 28227 Sugar Land, TX 77478 704-831-8950 - Charlotte Office/Direct 281-240-6500 - Main Switchboard 704-942-6773 - Mobile

Appendix A Document 3960C, Rev. 0, Attachment 4 AppendixAWBNU1ScopeandDrawingList Page 45 of 66 PreparerBy:JoseOrtiz ReviewedBy:CharlesDriskell Group ComponentId CompDescription ActuatorType ValveType Manufacturer ValveDwg/VendorManual 261 1CKV261260 REACTORBLDGHPFPSUPPLYHDRCHECK SA CK BORGWARNERCORP. 421JBB1002 261 1CKV261296 REACTORCOOLANTPUMPSPRINKLERHDRISOLCHK SA CK BORGWARNERCORP. 421JBB1002 262 1FCV26240 REACTORBLDGSTANDPIPEISOL MO GA ANCHORDARLING 9314978 262 1FCV26243 REACTORCOOLANTPUMPSPRINKLERHDRISOL MO GA ANCHORDARLING 9314978 321 1BYV32288 CONTROLAIR1FCV32110BYPASS M GL K085/KEROTEST 99099X01S 321 1BYV32298 ESSENTCONTROLAIR1FCV3280BYPASS M GL K085/KEROTEST 99099X01S 321 1BYV32308 ESSENTCONTROLAIR1FCV32102BYPASS M GL K085/KEROTEST 99099X01S REVISE 322 1CKV32293 CONTROLAIRCNTMTCHECK SA CK K085/KEROTEST 2 UNID ds322 1CKV32303 ESSENTCNTLAIRCNTMTCHECK SA CK K085/KEROTEST drawing dsy 322 1CKV32313 ESSENTCNTLAIRCNTMTCHECK SA CK K085/KEROTEST nos. ds323 1FCV3280 ESSENTCONTROLAIRTRACNTMTISOL AO GL L170/LESLIECO. 717543070D 323 1FCV32102 ESSENTCONTROLAIRTRBCNTMTISOL AO GL L170/LESLIECO. 717543070D 323 1FCV32110 CONTROLAIRCNTMTISOL AO GL L170/LESLIECO. 717543070D 431 1CKV43834 PASWASTETOCNTMTSUMPCHECK SA CK CircleSeals N89180 431 1CKV43841 PASWASTETOCNTMTSUMPCHECK SA CK CircleSeals N89180 431 1CKV43883 PASCONTAINMENTAIRRETURNCHECK SA CK CircleSeals N89180 431 1CKV43884 PASCONTAINMENTAIRRETURNCHECK SA CK CircleSeals N89180 433 1FCV43202 LOCAH2CNTMTMONITOROUTLETISOL SO GL TargetRock 101500532 433 1FCV43208 LOCAH2CNTMTMONITOROUTLETISOL SO GL TargetRock 101500532 433 1FCV43434 LOCAH2CNTMTMONITORD/SSAMPLEISOL SO GL TargetRock 101500532 433 1FCV43436 LOCAH2CNTMTMONITORD/SSAMPLEISOL SO GL TargetRock 101500532 433 1FSV43307 PASCONTAINMENTAIRRETURNISOL SO GL TargetRock 101500532 433 1FSV43325 PASCONTAINMENTAIRRETURNISOL SO GL TargetRock 101500532 521 1ISV52500 PENETRATION26BILRTOUTSIDE M GL DragonValve 13824 521 1ISV52501 PENETRATION26AILRTOUTSIDE M GL DragonValve 13824 521 1ISV52502 PEN96AINTERGRATEDLEAKRATETESTOUTSIDE M GL DragonValve 13824 521 1ISV52503 PEN96BINTERGRATEDLEAKRATETESTOUTSIDE M GL DragonValve 13824 521 1ISV52504 PENETRATION26BILRTINSIDE M GL DragonValve 13824 521 1ISV52505 PENETRATION26AILRTINSIDE M GL DragonValve 13824 521 1ISV52506 PEN96AINTERGRATEDLEAKRATETESTINSIDE M GL DragonValve 13824 521 1ISV52507 PEN96BINTERGRATEDLEAKRATETESTINSIDE M GL DragonValve 13824 611 1CKV61533 GLYCOLSUPPLYHEADERBYPASSCHECK SA CK A391/ANCHORDARLING W9825144 611 1CKV61680 GLYCOLRETURNHEADERBYPASSCHECK SA CK A391/ANCHORDARLING W9825144 611 1CKV61692 GLYCOLCOOLEDFLOORSUPPLYBYPASSCHECK SA CK A391/ANCHORDARLING W9825144 611 1CKV61745 GLYCOLCOOLEDFLOORRETURNBYPASSCHECK SA CK A391/ANCHORDARLING W9825144 623 1FCV6261 CVCSSEALWATERRETURNHEADERISOL MO GA W120/WESTINGHOUSEELECW120/WESTINGHOUSEELEC 115E001 623 1FCV6263 CVCSSEALWATERRETURNHEADERISOL MO GA W120/WESTINGHOUSEELECW120/WESTINGHOUSEELEC 115E001 624 1CKV62639 CVCSSEALWTR1FCV6261EQLCHECK SA CK K085/KEROTESTMANUFACTURINGCORP. 7500001295TVDD9556 Typical for group 32-2:add change paper DCA-38814-08 to all 3 drawings

Appendix A Document 3960C, Rev. 0, Attachment 4 AppendixAWBNU1ScopeandDrawingList Page 46 of 66 631 1CKV63868 1CKV63868CONTAINMENTN2HEADERCHECK SA CK K085/KEROTESTMANUFACTURINGCORP. TVW130608GS(2) 632 1FCV6364 SISACCUMN2HDRINLETVLV AO GL F130/FISHERCONTROLSCOINC 54A0240 633 1FCV6323 COLDLEGACCUMULATORFILLFROMSIP1AAISV AO GL F130/FISHERCONTROLSCOINC 54A0223 634 1FCV6371 SISCHECKVLVTESTLINEHOLDUPTANKISOL AO GL F130/FISHERCONTROLSCOINC 54A0237 635 1FCV6384 SISCHECKVLVLEAKTESTHOLDUPTANKISOL AO GL F130/FISHERCONTROLSCOINC 54A0237 671 1CKV67575A 1FCV6787BYPASSCHECK SA CK ANCHORDARLING W9825144 671 1CKV67575B 1FCV67103BYPASSCHECK SA CK ANCHORDARLING W9825144 671 1CKV67575C 1FCV6795BYPASSCHECK SA CK ANCHORDARLING W9825144 671 1CKV67575D 1FCV67111BYPASSCHECK SA CK ANCHORDARLING W9825144 672 1CKV67580A UPPERCNTMTVENTCLR1AERCWSUPHDRCHECK SA CK A585/ATWOOD&MORRILLCO 1473502 672 1CKV67580B UPPERCNTMTVENTCLR1BERCWSUPHDRCHECK SA CK A585/ATWOOD&MORRILLCO 1473502 672 1CKV67580C UPPERCNTMTVENTCLR1CERCWSUPHDRCHECK SA CK A585/ATWOOD&MORRILLCO 1473502 672 1CKV67580D UPPERCNTMTVENTCLR1DERCWSUPHDRCHECK SA CK A585/ATWOOD&MORRILLCO 1473502 673 1CKV67585A 1FCV67295BYPASSCHECK SA CK K085/KEROTEST 72576978 673 1CKV67585B 1FCV67297BYPASSCHECK SA CK K085/KEROTEST 72576978 673 1CKV67585C 1FCV67296BYPASSCHECK SA CK K085/KEROTEST 72576978 673 1CKV67585D 1FCV67298BYPASSCHECK SA CK K085/KEROTEST 72576978 675 1FCV67130 UPPERCNTMTVENTCLR1AERCWSUPHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67131 UPPERCNTMTVENTCLR1AERCWRETHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67133 UPPERCNTMTVENTCLR1CERCWSUPHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67134 UPPERCNTMTVENTCLR1CERCWRETHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67138 UPPERCNTMTVENTCLR1BERCWSUPHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67139 UPPERCNTMTVENTCLR1BERCWRETHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67141 UPPERCNTMTVENTCLR1DERCWSUPHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67142 UPPERCNTMTVENTCLR1DERCWRETHDRISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67295 UPPERCNTMTVENTCLR1AERCWRETISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67296 UPPERCNTMTVENTCLR1CERCWRETISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67297 UPPERCNTMTVENTCLR1BERCWRETISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 675 1FCV67298 UPPERCNTMTVENTCLR1DERCWRETISOL MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 681 1CKV68849 PRESSURIZERRELIEFTANKN2SUPHDRCHECK SA CK Kerotest 30508GLS(2) 683 1FCV68307 PRESSURIZERRELIEFTANKGASANALYZERSUPPLY AO GL C635/COPESVULCAN,INC. E171555 683 1FCV68308 PRESSURIZERRELIEFTANKGASANALYZERSUPPLY AO GL C635/COPESVULCAN,INC. E171555 701 1CKV70679 RCPTHERMALBARRIERCCSSUPHDRCHECK SA CK ATWOOD&MORRILL 1473501 703 1FCV7087 THERMALBARRIERCCSRETURN MO GA Walworth SA20991 703 1FCV70134 THERMALBARRIERCCSSUPPLY MO GA Walworth SA21173 704 1FCV7090 THERMALBARRIERCCSRETURN MO GA Walworth SA20991 772 1FCV77127 RBSUMPDISCHARGEFLOWCONTROL AO PLG Tufline/Xomox NP1211C 772 1FCV77128 RBSUMPDISCHARGEFLOWCONTROL AO PLG Tufline/Xomox NP1211C

Appendix A Document 3960C, Rev. 0, Attachment 4 AppendixAWBNU1ScopeandDrawingList Page 47 of 66 811 1CKV81502 PRIMARYWATERCNTMTHDRCHECKVLV SA CK W120/WESTINGHOUSEELECCORP 934D174 901 1FCV90110 CNTMTBLDGLOWERCOMPTAIRRADMONRETURN AO GL K085/KEROTEST TVD9957X01AC 901 1FCV90111 CNTMTBLDGLOWERCOMPTAIRRADMONRETURN AO GL K085/KEROTEST TVD9957X01AC 901 1FCV90116 CNTMTBLDGUPPERCOMPTAIRRADMONRETURN AO GL K085/KEROTEST TVD9957X01AC 901 1FCV90117 CNTMTBLDGUPPERCOMPTAIRRADMONRETURN AO GL K085/KEROTEST TVD9957X01AC Notes:1 tC11(seatcontact),C14(controlswitchtrip),andC16(final/maxseatingthrust).unningloadforspringtoclose.Indicatewhicharespringtoclose.Ifairtoclosetheminimumregulator/supplypressureisalsoneeded.

Appendix A Document 3960C, Rev. 0, Attachment 4 AttachmentBWBNU2ScopeandDrawingList Page 48 of 66 PreparerBy:JoseOrtiz ReviewedBy:CharlesDriskell Group ComponentId CompDescription SystemNo ActuatorType ValveType Manufacturer ValveDwg/VendorManual 261 2CKV261260 REACTORBLDGHPFPSUPPLYHDRCHECK 026 SA CK BORGWARNERCORP. 421JBB1002 261 2CKV261296 REACTORCOOLANTPUMPSPRINKLERHDRISOLCHK 026 SA CK BORGWARNERCORP. 421JBB1002 262 2FCV26240 REACTORBLDGSTANDPIPEISOL 026 MO GA ANCHORDARLING 9314978 262 2FCV26243 REACTORCOOLANTPUMPSPRINKLERHDRISOL 026 MO GA ANCHORDARLING 9314978 311 2CKV313378 INCOREINSTRRMAHU2BCWSLEAKRATECHECK 031 SA CK F990/FLOWSERVECORPORATION TVSW30604GS 311 2CKV313392 INCOREINSTRRMAHU2BCWRLEAKRATECHECK 031 SA CK F990/FLOWSERVECORPORATION TVSW30604GS 311 2CKV313407 INCOREINSTRRMAHU2ACWSLEAKRATECHECK 031 SA CK K085/KEROTESTMANUFACTURINGCORP. TVSW30604GS 311 2CKV313421 INCOREINSTRRMAHU2ACWRLEAKRATECHECK 031 SA CK K085/KEROTESTMANUFACTURINGCORP. TVSW30604GS 312 2FCV31305 INCOREINSTRRMAHU2ACWRISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31306 INCOREINSTRRMAHU2ACWRISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31308 INCOREINSTRRMAHU2ACWSISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31309 INCOREINSTRRMAHU2ACWSISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31326 INCOREINSTRRMAHU2BCWRISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31327 INCOREINSTRRMAHU2BCWRISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31329 INCOREINSTRRMAHU2BCWSISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 312 2FCV31330 INCOREINSTRRMAHU2BCWSISOL 031 AO GL T340/TUFLINEDIV/XOMOXCORP NP1071C 321 2BYV32318 ESSENTCONTROLAIR2FCV32103BYPASS 032 M GL K085/KEROTESTMANUFACTURINGCORP. TVD9909X01S(2) 321 2BYV32328 ESSENTCONTROLAIR2FCV3281BYPASS 032 M GL K085/KEROTESTMANUFACTURINGCORP. TVD9909X01S(2) 321 2BYV32338 CONTROLAIR2FCV32111BYPASS 032 M GL K085/KEROTESTMANUFACTURINGCORP. TVD9909X01S(2) 322 2CKV32323 ESSENTCNTLAIRCNTMTCHECK 032 SA CK K085/KEROTESTMANUFACTURINGCORP. dsDh) 322 2CKV32333 ESSENTCNTLAIRCNTMTCHECK 032 SA CK K085/KEROTESTMANUFACTURINGCORP. dsDh) 322 2CKV32343 CONTROLAIRCNTMTCHECK 032 SA CK K085/KEROTESTMANUFACTURINGCORP. dsDh 323 2FCV3281 ESSENTCONTROLAIRTRACNTMTISOL 032 AO GL L170/LESLIECO. 717543070D 323 2FCV32103 ESSENTCONTROLAIRTRBCNTMTISOL 032 AO GL L170/LESLIECO. 717543070D 323 2FCV32111 CONTROLAIRCNTMTISOL 032 AO GL L170/LESLIECO. 717543070D 433 2FCV43202 LOCAH2CNTMTMONITOROUTLETISOL 043 SO GL TargetRock 82KK003BBSH1&2 433 2FCV43434 LOCAH2CNTMTMONITORD/SSAMPLEISOL 043 SO GL TargetRock 82KK003BBSH1&2 521 2ISV52500 PENETRATION26BILRTOUTSIDE 052 M GA DragonValve 13824 521 2ISV52501 PENETRATION26AILRTOUTSIDE 052 M GA DragonValve 13824 521 2ISV52502 PEN96AINTERGRATEDLEAKRATETESTOUTSIDE 052 M GA DragonValve 13824 521 2ISV52503 PEN96BINTERGRATEDLEAKRATETESTOUTSIDE 052 M GA DragonValve 13824 521 2ISV52504 PENETRATION26BILRTINSIDE 052 M GA DragonValve 18437 521 2ISV52505 PENETRATION26AILRTINSIDE 052 M GA DragonValve 13824 521 2ISV52506 PEN96AINTERGRATEDLEAKRATETESTINSIDE 052 M GA DragonValve 13824 521 2ISV52507 PEN96BINTERGRATEDLEAKRATETESTINSIDE 052 M GA DragonValve 13824 611 2CKV61533 GLYCOLSUPPLYHEADERBYPASSCHECK 061 SA CK A391/ANCHORDARLING W9825144 611 2CKV61680 GLYCOLRETURNHEADERBYPASSCHECK 061 SA CK A391/ANCHORDARLING W9825144 611 2CKV61692 GLYCOLCOOLEDFLOORSUPPLYBYPASSCHECK 061 SA CK A391/ANCHORDARLING W9825144 611 2CKV61745 GLYCOLCOOLEDFLOORRETURNBYPASSCHECK 061 SA CK A391/ANCHORDARLING W9825144 REVISE 3 drawings to reflect U2 dedicated drawing for soft seat.

Appendix A Document 3960C, Rev. 0, Attachment 4 AttachmentBWBNU2ScopeandDrawingList Page 49 of 66 623 2FCV6261 CVCSSEALWATERRETURNHEADERISOL 062 MO GA W120/WESTINGHOUSEELECW120 115E001 623 2FCV6263 CVCSSEALWATERRETURNHEADERISOL 062 MO GA W120/WESTINGHOUSEELECW120 115E001 624 2CKV62639 CVCSSEALWTR2FCV6261EQLCHECK 062 SA CK K085/KEROTESTMANUFACTURINGCORP. TVDD9556 631 2CKV63868 2CKV63868CONTAINMENTN2HEADERCHECK 063 SA CK K085/KEROTESTMANUFACTURINGCORP. TVW130608GS(2) 632 2FCV6364 SISACCUMN2HDRINLETVLV 063 AO GL F130/FISHERCONTROLSCOINC 54A0240 633 2FCV6323 COLDLEGACCUMULATORFILLFROMSIP1AAISV 063 AO GL F130/FISHERCONTROLSCOINC 54A0223 634 2FCV6371 SISCHECKVLVTESTLINEHOLDUPTANKISOL 063 AO GL FISHERCONTROLSCO. 54A0237 634 2FCV6384 SISCHECKVLVLEAKTESTHOLDUPTANKISOL 063 AO GL FISHERCONTROLSCO. 54A0237 671 2CKV67575A 2FCV6787BYPASSCHECK 067 SA CK ANCHORDARLING W9825144 671 2CKV67575B 2FCV67103BYPASSCHECK 067 SA CK ANCHORDARLING W9825144 671 2CKV67575C 2FCV6795BYPASSCHECK 067 SA CK ANCHORDARLING W9825144 671 2CKV67575D 2FCV67111BYPASSCHECK 067 SA CK ANCHORDARLING W9825144 672 2CKV67580A UPPERCNTMTVENTCLR1AERCWSUPHDRCHECK 067 SA CK A585/ATWOOD&MORRILLCO 1473502 672 2CKV67580B UPPERCNTMTVENTCLR2BERCWSUPHDRCHECK 067 SA CK A585/ATWOOD&MORRILLCO 1473502 672 2CKV67580C UPPERCNTMTVENTCLR2CERCWSUPHDRCHECK 067 SA CK A585/ATWOOD&MORRILLCO 1473502 672 2CKV67580D UPPERCNTMTVENTCLR2DERCWSUPHDRCHECK 067 SA CK A585/ATWOOD&MORRILLCO 1473502 673 2CKV67585A 2FCV67295BYPASSCHECK 067 SA CK K085/KEROTEST 72576978 673 2CKV67585B 2FCV67297BYPASSCHECK 067 SA CK K085/KEROTEST 72576978 673 2CKV67585C 2FCV67296BYPASSCHECK 067 SA CK K085/KEROTEST 72576978 673 2CKV67585D 2FCV67298BYPASSCHECK 067 SA CK K085/KEROTEST 72576978 676 2FCV67130 UPPERCNTMTVENTCLR1AERCWSUPHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67131 UPPERCNTMTVENTCLR1AERCWRETHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67133 UPPERCNTMTVENTCLR1CERCWSUPHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67134 UPPERCNTMTVENTCLR1CERCWRETHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67138 UPPERCNTMTVENTCLR1BERCWSUPHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67139 UPPERCNTMTVENTCLR1BERCWRETHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67141 UPPERCNTMTVENTCLR1DERCWSUPHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67142 UPPERCNTMTVENTCLR1DERCWRETHDRISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67295 UPPERCNTMTVENTCLR2AERCWRETISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67296 UPPERCNTMTVENTCLR2CERCWRETISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67297 UPPERCNTMTVENTCLR2BERCWRETISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 676 2FCV67298 UPPERCNTMTVENTCLR2DERCWRETISOL 067 MO PLG T340/TUFLINEDIV/XOMOXCORP NP3491C 681 2CKV68849 PRESSURIZERRELIEFTANKN2SUPHDRCHECK 068 SA CK Kerotest 30508GLS(2) 683 2FCV68307 PRESSURIZERRELIEFTANKGASANALYZERSUPPLY 068 AO GL C635/COPESVULCAN,INC. E171555 683 2FCV68308 PRESSURIZERRELIEFTANKGASANALYZERSUPPLY 068 AO GL C635/COPESVULCAN,INC. E171555 701 2CKV70679 RCPTHERMALBARRIERCCSSUPHDRCHECK 070 SA CK Flowserve 13103681001 704 2FCV7087 THERMALBARRIERCCSRETURN 070 MO GA Walworth SA20991 705 2FCV70134 THERMALBARRIERCCSSUPPLY 070 MO GA Walworth SA21173

Appendix A Document 3960C, Rev. 0, Attachment 4 AttachmentBWBNU2ScopeandDrawingList Page 50 of 66 706 2FCV7090 THERMALBARRIERCCSRETURN 070 MO GA Walworth 095710501 772 2FCV77127 RBSUMPDISCHARGEFLOWCONTROL 077 AO PLG Tufline/Xomox NP1211C 772 2FCV77128 RBSUMPDISCHARGEFLOWCONTROL 077 AO PLG Tufline/Xomox NP1211C 811 2CKV81502 PRIMARYWATERCNTMTHDRCHECKVLV 081 SA CK W120/WESTINGHOUSEELECCORP 934D174 901 2FCV90110 CNTMTBLDGLOWERCOMPTAIRRADMONRETURN 090 AO GL K085/KEROTEST 105962101sheets1,2,3 901 2FCV90111 CNTMTBLDGLOWERCOMPTAIRRADMONRETURN 090 AO GL K085/KEROTEST 105962101sheets1,2,3 901 2FCV90116 CNTMTBLDGUPPERCOMPTAIRRADMONRETURN 090 AO GL K085/KEROTEST 105962101sheets1,2,3 901 2FCV90117 CNTMTBLDGUPPERCOMPTAIRRADMONRETURN 090 AO GL K085/KEROTEST 105962101sheets1,2,3

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 51 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page A52 of A66 Group 32-2 Document 3960C, Rev. 0, Attachment 4 Appendix A Page 52 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 53 of 66 Manual No. 800-PC Issued: March 31, 2004 INSTRUCTION MANUAL for 1/2" thru 2" 800 lb. Piston Lift Check Valves with Resilient Seat Option Flowserve Corporation Flow Control Division 1900 S. Saunders Street P.O. Box 1961 Raleigh, NC 27603 Phone: (919) 832-0525 FAX: (919) 831-3369

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 54 of 66 Table of Contents 1.0 Physical Description and Operation of Equipment 2.0 Design Conditions 3.0 Operating Conditions 4.0 Test Conditions 5.0 Operating Precautions and Limitations 6.0 Installation Instructions 7.0 Maintenance Requirements 8.0 Periodic Inservice Testing Recommendations and Procedures 9.0 Maintenance Instructions 10.0 Storage Requirements 11.0 Reference Drawings

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 55 of 66 Manual No. 800-PC Revision Sheet Revision Date Changes- 03/31/2004 Original Issue

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 56 of 66 1.0 PHYSICAL DESCRIPTION AND OPERATION OF EQUIPMENT 1.1 Piston Check Valves (Figures 1 & 2 - See Section 11.0)Piston Lift Check valves are generally used in applications where pressure drop through the valve is not critical, although Flowserve piston check valves have a relatively low pressure drop.These small piston lift check valves include a return spring to facilitate closing.All Flowserve piston check valves have body guided discs to provide resistance to wear, thus, insuring a longer life. An equalizing hole is provided in the disc as a drain for condensate in steam valves.A dual seat may be supplied which provides a hard surface for high differential pressure sealing and a resilient seat for sealing during low differential pressure.These instructions are being furnished to the customer for use in the installation, operation and maintenance of the 800 pressure class series piston lift check valves.2.0 DESIGN CONDITIONS N/A 3.0 OPERATING CONDITIONS N/A 4.0 TEST CONDITIONS 4.1 Each valve covered by this manual has received the following hydrostatic tests:4.1.1 Shell hydrostatic test at 1.5 times the 100°F pressure rating.4.1.2 A seat leakage and disc closure test at 110% of the 100°F pressure rating.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 57 of 66 4.0 TEST CONDITIONS (Continued) 4.2 Each valve, supplied with resilient seated discs, received the following hydrostatic tests:4.2.1 A low-pressure water seat test at 50 psig for containment isolation valves.5.0 OPERATING PRECAUTIONS AND LIMITATIONS 5.1 Maximum hydrostatic test pressure shall not exceed the values imposed by the ASME Code, Section III.6.0 INSTALLATION INSTRUCTIONS 6.1 Lifting and Handling Requirements and Limitations 6.1.1 Good judgement should be exercised in selecting a lifting device that will safely support the unit's weight.6.1.2 Remove the end covers.6.1.3 Remove any blocks or heavy paper that might have been used to keep the disc from moving during shipment.6.2 Installation 6.2.1 Although the valves have been shipped in a clean condition, prior to installing the valves, examine the lines and the valve ports for foreign matter and clean them thoroughly if they have been exposed to the elements. (BEFORE CLEANING IN THIS FASHION, CHECK AT THE SITE TO SEE IF A SPECIFIC CLEANING PROCEDURE SHOULD BE FOLLOWED.) Flush the valves out with water if possible; otherwise blow them out with air or steam.In performing this cleaning procedure, the ports should be vertical to assure complete removal of all matter which might have accumulated during storage.6.2.2 Ensure that there is no line sag at the point of installation. Eliminate any pipeline deviation by the proper use of pipeline hangers or similar device.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 58 of 66 6.2 Installation (Continued) 6.2.3 Extreme caution should be taken when installing check valves. The arrow on the valve body indicates flow direction. Therefore, when installing a check valve, place it so that the flow of the incoming fluid will open the valve and return flow will close it. Check valves installed in reverse position will stop the flow in the normal flow direction. Valves should be installed in a horizontal run of pipe with the gasket retainer on top.Maximum deviation from the horizontal should be +/- 15°.6.2.4 The valves should then be blocked or slung into position with apparatus that is sufficient to hold the valve assembly weight while the valve is being welded into the line. WELDING SHOULD TAKE PLACE WITH THE DISC IN THE OPEN POSITION. This is particularly important for valves with soft seats. Welding the valve with the disc closed will damage the resilient seat material. This may require removal of the internals prior to welding.6.2.5 Remove the end protectors and clean the ends with a solvent such as acetone in preparation to welding.7.0 MAINTENANCE REQUIREMENTS 7.1 Preventative Maintenance 7.1.1 Check all bolts periodically to ensure tightness and to forestall possible leaks.7.2 Recommended Spare Parts 7.2.1 Recommended spare parts are pressure seal gasket (030), bonnet (002),disc assembly (004 & 005, 245 & 306 if equipped with resilient seated disc) and spring (429). The recommended quantity is 1 set for every 10 valves.7.2.2 For consolidating spare parts (See 7.2.1), use the following guidelines:

~~- 1/2" thru 1" Piston Check Valves Recommended spare parts are interchangeable throughout this size range.~~

Note that similar materials should be ordered for valve body type (i.e.carbon steel bonnet for carbon steel valve).

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 59 of 66 7.0 MAINTENANCE REQUIREMENTS (Continued)

~~- 11/2" and 2" Reduced Port Piston Check Valves Recommended spare parts are interchangeable throughout this size range.~~

Note that similar materials should be ordered for valve body type (i.e.carbon steel bonnet for carbon steel valve).

~~- 2" Full Port Piston Check Valves Recommended spare parts are only interchangeable with other 2" Full Port Piston Check Valves. Same material restrictions stated above apply.~~

7.3 Lubrication 7.2.1 A light coating of lubricant should also be applied to the bonnet retainer threads if and when the valve is reassembled.8.0 PERIODIC INSERVICE TESTING RECOMMENDATIONS AND PROCEDURES 8.1 This is not required for piston check valves without external operators.9.0 MAINTENANCE INSTRUCTIONS 9.1 Disassembly WARNING PRIOR TO PERFORMING DISASSEMBLY, CLOSE OFF THE LINE PRESSURE TO THE VALVE, AND RELEASE ALL PRESSURE IN THE VALVE.9.1.1 Remove the anti-rotation pin (258) and the bonnet capscrew (216). The gasket retainer (033) may now be unscrewed and removed. Now thread the bonnet capscrew (216) directly into bonnet (002) and pull bonnet capscrew (216) directly upward.Care should be taken to pull evenly and straight upward as not to score the neck walls of the valve and bonnet edges.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 60 of 66 9.0 MAINTENANCE INSTRUCTIONS (Continued)Pulling of the capscrew (216) will remove the bonnet (002) and pressure seal gasket (030). Lift out the spring (429) and disc (004).It may be necessary to insert a wire hook into the holes located in the side of the disc in order to lift the disc out of the valve.9.1.2 After removal of the disc from the valve, care should be taken to protect the seating surface from damage. The disc should be placed in a clean area until it is ready for replacement. THE SLIGHTEST NICK OR SCRATCH ON SEATING SURFACE MAY PREVENT COMPLETE SHUTOFF AND NECESSITATE EXTENSIVE REWORK OR REPLACEMENT.9.1.3 Resilient Seat Removal (if so equipped)To disassemble disc/resilient seat assembly:a) Remove retaining ring (245) b) Unscrew disc (004) from disc skirt (005)Note: For removal, disc skirt should be held by the relieved outside diameter containing the drainage hole rather than by the outside diameter guiding surfaces. A slot for a fitted screw driver is provided at the top of the disc to facilitate removal and to prevent damage to these critical surfaces.c) Remove the resilient seat by carefully slipping it over the disc.CAUTION DO NOT EXPOSE THE RESILIENT SEAT TO ANY PETROLEUM BASED OILS OR GREASES OR OTHER CUTTING FLUIDS, LUBRICANTS, ETC. WHICH ARE HYDROCARBON BASED.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 61 of 66 9.0 MAINTENANCE INSTRUCTIONS (Continued) 9.2 Refinishing Sealing Surfaces Minor discontinuities in the seat sealing surface, which may cause leakage can, in many cases, be removed by lapping. Major defects such as cracks or deep gouges will generally require replacement of the part.Minor discontinuities on the valve disc sealing surfaces may be removed by remachining the surface to remove a few thousandths of material. Major defects will generally require replacement of the part.(NOTE: Lapping is a polishing process where a sealing surface is ground with an abrasive held in place by a special fixture. The abrasive is commonly found in paste form or bonded to a paper backing.Detailed instructions on the use of lapping abrasives and fixtures, normally supplied with such equipment, should be adhered to.)In order to maintain seat tightness in piston check valves, the sealing surfaces on both the disc and seat ring must be kept within close tolerances. Flowserve does not recommend lapping the disc directly to the seat. A good seal is dependent on line contact. Direct contact lapping will result in excessive seat widths.Lapping equipment for the series 800 piston lift check valve seat is available through Flowserve. Contact your nearest Flowserve representative for information.9.3 Reassembly 9.3.1 First, all dirt, scale and foreign matter should be removed from inside the valve body and bonnet.9.3.2 Before reassembling the valve, check the seating surfaces to determine that no scratches or minor imperfections are on the disc or seat ring. If any are evident - lap these surfaces until none are visible. (Reference Para. 9.2)

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 62 of 66 9.0 MAINTENANCE INSTRUCTIONS (Continued) 9.3.3 Resilient Seat Assembly (if so equipped) a) Carefully place the resilient seat over the disc.b) Reassembly of the balance of the assembly is the reverse of that described in 9.1.3.c) Use nuclear grade thread lubricant on the threads, making sure that it does not come in contact with the resilient seat material.d) The disc skirt (005) is to be screwed firmly against the disc (004) shoulder prior to installing the retaining ring (245).9.3.4 Reassembly of the valve is accomplished by inserting the disc or disc assembly (004) and spring (429), followed by the bonnet (002) and pressure seal gasket (030). Use a nuclear grade thread lubricant on the threads, making sure that it does not come in contact with the resilient seat material. Then screw in the gasket retainer (033). The bonnet capscrew (216) is then threaded into the bonnet through the hole in the retainer (033). A maximum of 5 ft-lbs of torque should be exerted on the bonnet capscrew. Insert the anti-rotation pin (258) in hole in top of the body (001).NOTE: Consolidation of graphite during initial system pressurization is normal and will often cause the bonnet capscrew to become finger tight or even loose.9.3.5 Retorque the bonnet capscrew to 50 to 60 inch-lbs. when the system is initially pressurized to ensure that the bonnets will not move out of the sealed position when the pressure is relieved.9.4 Trouble Shooting A. Leakage Between the Disc (004) and Seat Ring (013)This could be an indication that there is foreign matter on the seating surfaces.Disassemble the valve and remove the source of the trouble. If no foreign matter is found, inspect the seating surfaces of the valve for signs of a scarred or damaged seat - in which case the seating surfaces of the Disc (004) and Seat Ring (013) should be lapped until no visible defects remain. (Refer to Para. 9.2)

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 63 of 66 10.0 STORAGE REQUIREMENTS The valves have been shipped in the partially open position. Upon receipt of the valves at destination, the crates should be examined thoroughly for signs of mishandling or damage during shipment. With the valves strapped to the shipping skids, all bolting should be checked to ensure that the joints are secure. Bolting on occasion, may become loosened during shipment and handling.The valves should then be stored in a sheltered area to protect them from the elements, dirt and foreign material. They should not be exposed to the atmosphere, uncrated or removed from the shipping skids except in a clean area just prior to installation.If the valves are not to be installed within a short period of time after receipt, and will require long-term storage, the following should be adhered to:(a) They should be stored in an upright position and where there is minimal temperature variations and the temperature does not drop below 50°F.(b) In their storage condition, the valves should be wrapped in polyethylene to prevent accumulation of dust or foreign matter.(c) A check-off tag should be affixed to each unit and should be dated and signed off by the inspector witnessing the inspection which is recommended at 6-month intervals.The shelf life for resilient seat materials is 5 years.The shelf life for gaskets is indefinite.

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 64 of 66 SECTION 11.0 REFERENCE DRAWINGS

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 65 of 66

Appendix A Document 3960C, Rev. 0, Attachment 4 Page 66 of 66

Document 3960C, Rev. 0, Attachment 5 Page 2 Revisions Rev. DCR/N Pages No. No. Description of Changes Affected 0 N/A Initial release All Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 3 Table of Contents Page 1 OBJECTIVE AND SCOPE 5 1.1 Objective 5 1.2 Scope 5 1.3 Historical Leakage 6 2 METHODOLOGY 7 2.1 Variables 7 2.2 Globe Valve Sealing Load 8 2.2.1 Static Sealing Force Acting in Stem Axis Direction 9 2.2.2 Dynamic Sealing Force Acting in Stem Axis Direction due to Differential Pressure 9 2.2.3 Total Sealing Force and Seal Force Reduction 9 3 INPUTS 12 3.1 Calculation Inputs 12 4 ASSUMPTIONS 14 5 RESULTS, RECOMMENDATIONS, AND CONCLUSIONS 15 5.1 Seat Load Results 15 5.2 Seat Load Reduction and Maximum LLRT Test Pressure for 10% Seal Load Reduction 16 5.3 Effect of Seat Load Reduction on Seat Leakage for Metal Seated Valves 16 5.4 Notes/ Recommendations 18 5.5 Conclusion 18 6 REFERENCES 19 Appendix A - Supporting Documents Appendix B - Sample Calculations Pages Rev Main Text 20 0 Appendix A 14 0 Appendix B 03 0 Total 37 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 4 List of Tables Table Description Page Table 1-1: Analysis Scope 5 Table 1-2: Unit 1 and Unit 2 LLRT Leakage History 6 Table 3-1: Valve-Specific Input Data [8, 10] 12 Table 5-1: Seat Load Results 15 Table 5-2: Seat Load Reduction and Maximum DPtest for 10% Reduction in Seat Load 16 Table 5-3: Percentage Increase in Leakage Flow Area Calculation for Group 32-3 Valves 17 List of Figures Figure Description Page Figure 2-1: Force Components at Closed Position with Fluid Pressure Acting Above the Disc (Static and Dynamic) 8 Figure 2-2: Total Sealing Force Acting in the Stem Axis Direction (FS) and Normal to the Seat (Rr) as mentioned in Reference 5 10 Figure 5-1: Microscopic Flow Path Under Light and Heavy Seating Load [6] 18 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 5 1OBJECTIVE AND SCOPE 1.1 OBJECTIVE Kalsi Engineering, Inc. (KEI) has been contracted by Tennessee Valley Authority (TVA) to provide engineering services to evaluate the impact of local leak rate test (LLRT) pressures greater than the calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA), Pa, for cases where greater test pressure tends to increase the sealing force. This work is being done in accordance with Purchase Order No. 6232543 [2]1.The objective of this report is to determine the impact of the reduced LLRT pressure from DPtest to Pa on the seat leakage. All work performed under this project was done in accordance with the requirements of the KEI Quality Assurance Program [1], which meets the intent of 10CFR50 Appendix B requirements.1.2 SCOPE The scope of this attachment is LLRT air-operated globe valves. Component IDs and basic information are shown below in Table 1-1. All the valves are direct acting (extend to close) with a reverse acting (air to retract) actuator [7].Table 1-1: Analysis Scope Manufacturer/Group Component Id Comp Description Drawing No.1-FCV-32-80 1-FCV-32-102 ESSENT CONTROL AIR TR A and 1-FCV-32-110 L170/LESLIE CO./32-3 B CNTMT ISOL; CONTROL AIR 2-FCV-32-81 717543070D CNTMT ISOL 2-FCV-32-103 2-FCV-32-111 F130/FISHER 63-2 1,2-FCV-63-64 SIS ACCUM N2 HDR INLET VLV CONTROLS CO INC/54A0240 1The number in [] shows the reference documented in Section 6.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 6 Manufacturer/Group Component Id Comp Description Drawing No.F130/FISHER COLD LEG ACCUMULATOR FILL 63-3 1,2-FCV-63-23 CONTROLS CO INC/FROM SIP 1A-A ISV 54A0223 F130/FISHER SIS CHECK VLV TEST LINE 63-4 1,2-FCV-63-71 CONTROLS CO INC/HOLDUP TANK ISOL 54A0237 F130/FISHER SIS CHECK VLV LEAK TEST 63-5 1,2-FCV-63-84 CONTROLS CO INC/HOLDUP TANK ISOL 54A0237 1,2-FCV-68-307 PRESSURIZER RELIEF TANK C635/COPES-VULCAN, 68-3 1,2-FCV-68-308 GAS ANALYZER SUPPLY INC./ E171555 1.3 HISTORICAL LEAKAGE Reference 3 and Reference 8 provide the LLRT history for the Unit 1 and Unit 2 AOVs. Table 1-2, below, summarizes the results.Table 1-2: Unit 1 and Unit 2 LLRT Leakage History Group Component Id Leakage Results Seat Type 1-FCV-32-80 Favorable history 1-FCV-32-102 Favorable history 1-FCV-32-110 Favorable history 32-3 Hard 2-FCV-32-81 Favorable history 2-FCV-32-103 Favorable history 2-FCV-32-111 Unfavorable history 1-FCV-63-64 Favorable history 63-2 Hard 2-FCV-63-64 Favorable history 1-FCV-63-23 Unfavorable history 63-3 Hard 2-FCV-63-23 Favorable history 1-FCV-63-71 Favorable history 63-4 Hard 2-FCV-63-71 Favorable history 1-FCV-63-84 Unfavorable history 63-5 Hard 2-FCV-63-84 Favorable history 1-FCV-68-307 Favorable history 2-FCV-68-307 Favorable history 68-3 Hard 1-FCV-68-308 Favorable history 2-FCV-68-308 Favorable history Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 7 2METHODOLOGY 2.1 VARIABLES Variable Description Units Ao Area based on mean seat diameter = In2 ARLB Lower benchset diaphragm area (EDA) In2 As Stem area at packing = In2 Dm Mean seat diameter In DP Valve differential pressure Psi DPtest Bounding LLRT test differential pressure Psi Test differential that will result in a 10% reduction in sealing load at DPtest_10% Psi Pa ds Stem diameter at packing In FDF Disc-to-cage/body friction force Lb FDP_DPtest DP force acting on sealing diameter @ DPtest Lb FDP_Pa DP force acting on sealing diameter @ Pa Lb FDyn_DPtest Dynamic sealing force in stem axis direction @ DPtest Lb FDyn_Pa Dynamic sealing force in stem axis direction @ Pa Lb FPack Packing friction force Lb Maximum closed static friction load (includes packing and static seal FR Lb friction)FStatic Static sealing force in stem axis direction Lb FSF Static seal friction Lb FS_DP Sealing load due to differential pressure Lb FS_DPtest Total sealing force in stem axis direction @ DPtest Lb FS_Pa Total sealing force in stem axis direction @ Pa Lb FSPL Minimum spring preload Lb FW Sealing force due to disc and stem weight Lb Calculated peak containment internal pressure related to the design-Pa Psig basis loss-of-coolant accident (LOCA)Percentage reduction in sealing force due to the difference between Pa R %and DPtest PLB Minimum lower benchset Psig Rr_DPtest Sealing force normal to the seat @ DPtest Lb Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 8 Rr_DPtest_Lin Sealing force per linear inch normal to the seat @ DPtest Lb/in Rr_Pa Sealing force normal to the seat @ Pa Lb Rr_Pa_Lin Sealing force per linear inch normal to the seat @ Pa Lb/in Reduction in normal sealing force due to the difference between Rr_reduction Lb pressures DPtest and Pa Reduction in normal sealing force per linear inch due to the Rr_reduction_Lin Lb/in difference between pressures DPtest and Pa W Disc and stem weight Lb Stem angle from vertical Deg.Seat angle from stem axis Deg.Seat to disc friction coefficient 2.2 GLOBE VALVE SEALING LOAD During LLRT, sealing load includes 1) static force components and 2) dynamic force components due to differential pressure, DP. The subject valves are direct acting (extend to close) with a reverse acting (air to retract) actuator [7]. Therefore, at the closed position, the static force acting on the disc is equivalent to the spring preload minus packing/seal friction force. The dynamic force is proportional to the DP and the area over which the DP acts.The approach for this analysis is to:

1. Determine the reduction in total sealing force due to changing the test DP from the current value, DPtest, to the calculated peak pressure, Pa.
2. Determine the value of DPtest required to achieve a 10% reduction in sealing force at Pa.

Figure 2-1: Force Components at Closed Position with Fluid Pressure Acting Above the Disc (Static and Dynamic)Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 9 2.2.1 Static Sealing Force Acting in Stem Axis Direction The subject valves are direct acting (extend to close) with a reverse acting (air to retract) actuator[7]. Therefore, at the closed position, the static force, FStatic, acting on the disc is equivalent to the summation of spring preload and disc/stem weight minus packing/seal friction.

~~= + (1)~~

Where, = , and

~~= !~~

TVA has provided the maximum closed static friction load, FR, which includes packing and any other static friction. These valves do not have a seal between the disc and body/cage [10], therefore, FSF is zero.

= + " (2) 2.2.2 Dynamic Sealing Force Acting in Stem Axis Direction due to Differential Pressure The dynamic sealing force includes differential pressure force and disc-to-cage/body friction force: #$% = # # (3)

The DP sealing force at the closed position is:

~~# =( ' ) (4)~~

The disc-to-cage/body friction force, FDF in Equation 3 is zero at the closed position.2.2.3 Total Sealing Force and Seal Force Reduction The total sealing force acting in the stem axis direction is given by Equation 5:

~~= + #$% (5)~~

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 10 Figure 2-2: Total Sealing Force Acting in the Stem Axis Direction (FS) and Normal to the Seat (Rr) as mentioned in Reference 5 The total sealing force acting normal to the seat is given by Equation 6:

~~= (6) )~~

(sin - + cos -)The reduction in sealing force due to an LLRT test DP which is at a higher DP, DPtest, than the maximum peak pressure, Pa, is as follows:

~~)_)234 '% = )_# 2 )_ 5 (7)~~

Therefore, the percentage reduction in sealing force due to testing at DPtest which is higher than Pa is determined as follows:

~~)_)234 '% = 100 8 9 (8) )_# 2 Solving Equation 8 to determine the maximum allowable LLRT DPtest such that there is no greater than a 10% reduction is determined by the following equation:~~

0.1 + ( ' )

~~= (9) 2 _:;%~~

0.9 ( ' )Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 11 For each AOV in the scope of this analysis, the percentage reduction in sealing force, R, is determined using Equation 8 and the maximum LLRT test DP that results in a 10% reduction in sealing force is determined using Equation 9. A sample calculation is shown in Appendix B.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 12 3INPUTS 3.1 CALCULATION INPUTS The following inputs are used in the sealing load calculations:

1. The input data for the analyses are documented in Table 3-1 were obtained from calculations, drawings, specifications, and other information provided by TVA. Some of these inputs are provided in Appendix A.
2. The maximum permissible LLRT test pressure, DPtest, is 1.1 x 15 = 16.5 psig [9, 11].

Calculated peak containment internal pressure related to the design-basis loss-of-coolant accident (LOCA), Pa, for Watts Bar is 9.36 psig [9]. For purposes of this analysis, a lower and more conservative value of 9 psig is used.

~~3. The disc and stem weight, W, will be negligible compared to the spring preload and therefore are excluded from the static sealing load component calculated in Equation 2.~~

Since the weight term will generally provide additional closing force, this is a conservative.

4. The disc-to-seat coefficient of friction, µ, of 0.5 is used in Equation 7. The COF value does not affect the calculation of R (Equation 9) and DPtest_10% (Equation 10).
5. Justified assumptions were made where data were not available. It is important to note that the results of this analysis may be significantly affected by changing key inputs. It will be necessary to perform an impact analysis if key data are changed in the future.

Table 3-1: Valve-Specific Input Data [8, 10]Seat-to- Input Parameter Grou Component Disc PLB, ARLB, p ID ID FR, lbs Dm2, in ds, in 3, deg.Material psi in2 1-FCV-32-80/5.5 54.0 207 2.25 0.50 30 2-FCV-32-81 1-FCV-32-102/ Not 32-3 5.5 54.0 207 2.25 0.50 30 2-FCV-32-103 Available 1-FCV-32-110/5.5 54.0 207 2.25 0.50 30 2-FCV-32-111 2Dm values for Groups 63-2, 63-3, 63-4, 63-5, and 68-3 are based on assumption provided in Section 4.3 values are scaled from the drawings.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 13 Seat-to- Input Parameter Grou Component Disc PLB, ARLB, p ID ID FR, lbs Dm2, in ds, in 3, deg.Material psi in2 1-FCV-63-64/ Stellite-6/63-2 20.0 69.0 293 1.00 0.50 30 2-FCV-63-64 Stellite-6 1-FCV-63-23/ Stellite-6/63-3 23.0 69.0 453 1.00 0.50 30 2-FCV-63-23 316 SS 1-FCV-63-71/63-4 23.0 69.0 240 0.75 0.50 30 2-FCV-63-71 Stellite-6/1-FCV-63-84/ Stellite-6 63-5 23.0 69.0 410 0.75 0.50 30 2-FCV-63-84 1-FCV-68-307/16.0 17.0 111 0.38 0.31 30 2-FCV-68-307 316 SS/68-3 1-FCV-68-308/ 316 SS 16.0 17.0 111 0.38 0.31 30 2-FCV-68-308 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 14 4ASSUMPTIONS Data that have not been formally verified are treated as assumptions. Where possible, the basis of the data has been noted. The following general assumptions were used in this analysis.

1. The mean seat diameters, Dm, for Groups 63-2, 63-3, 63-4, 63-5, and 68-3 are not available and are assumed to be equal to the nominal valve size. The higher mean seat diameter increases the percentage reduction in the sealing force at the design basis accident pressure, Pa. A sensitivity analysis showed that increasing the mean seat diameter by 20% has a negligible increase the percentage reduction in the sealing force. This assumption does not require a verification.
2. The seat contact band width, t, for Group 32-3 valves is assumed to be of 0.005 inches (see Section 5.3) which is a reasonable assumption for a globe valve. A lower seat contact band width will provide a conservative results. A sensitivity analysis showed that lowering the seat contact band width to 0.001 inches does not change the overall conclusion because the percentage increase in the leakage area, AI, remains negligible. Therefore, this assumption does not require a verification.

Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 15 5RESULTS, RECOMMENDATIONS, AND CONCLUSIONS 5.1 SEAT LOAD RESULTS The sealing load per linear inch for the applicable AOV Globe valves are shown in Table 5-1. The minimum recommended seat contact force value for metal seats is 100 lb/in per Reference 6. The seat contact force values exceed 100 lb/in except for Group 32-3 valves at DPtest and Pa.Table 5-1: Seat Load Results Group Rr_DPtest_Lin, Rr_Pa_Lin, Rr_reduction_Lin, Component ID ID lb/in lb/in lb/in 1-FCV-32-80/23.10 18.80 4.30 2-FCV-32-81 1-FCV-32-102/32-3 23.10 18.80 4.30 2-FCV-32-103 1-FCV-32-110/23.10 18.80 4.30 2-FCV-32-111 1-FCV-63-64/63-2 374.16 372.65 1.51 2-FCV-63-64 1-FCV-63-23/63-3 390.20 388.69 1.51 2-FCV-63-23 1-FCV-63-71/63-4 614.57 613.73 0.84 2-FCV-63-71 1-FCV-63-84/63-5 537.24 536.40 0.84 2-FCV-63-84 1-FCV-68-307/146.98 146.75 0.23 2-FCV-68-307 68-3 1-FCV-68-308/146.98 146.75 0.23 2-FCV-68-308 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 16 5.2 SEAT LOAD REDUCTION AND MAXIMUM LLRT TEST PRESSURE FOR 10%SEAL LOAD REDUCTION The sealing force percentage reduction due to a decrease in LLRT pressure from DPtest to Pa (16.5 psig to 9 psig) is shown in Table 5-2. Table 5-2 also includes the maximum LLRT test pressure, DPtest_10%, that ensures no greater than a 10% seal load reduction at Pa.As can be seen that the seat load reduction for all the valves except for Group 32-3 are below 1%.The seat load is reduced by 18.61% for Group 32-3 valves when the pressure is reduced from DPtest to Pa (16.5 psig to 9 psig). The maximum test DP that results in 10% seat load reduction for Group 32-3 is 12.65 psig. Section 5.3 further discusses the effect of seat load reduction on seat leakage for Group 32-3 valves.Table 5-2: Seat Load Reduction and Maximum DPtest for 10% Reduction in Seat Load Component Rr_DPtest_Lin, Rr_reduction_Lin, DPtest_10%,Group ID R ID lb/in lb/in psig 1-FCV-32-80/23.10 4.30 18.61 12.65 2-FCV-32-81 1-FCV-32-102/32-3 23.10 4.30 18.61 12.65 2-FCV-32-103 1-FCV-32-110/23.10 4.30 18.61 12.65 2-FCV-32-111 1-FCV-63-64/63-2 374.16 1.51 0.40 215.04 2-FCV-63-64 1-FCV-63-23/63-3 390.20 1.51 0.39 223.90 2-FCV-63-23 1-FCV-63-71/63-4 614.57 0.84 0.14 619.80 2-FCV-63-71 1-FCV-63-84/63-5 537.24 0.84 0.16 542.84 2-FCV-63-84 1-FCV-68-307/146.98 0.23 0.16 540.08 2-FCV-68-307 68-3 1-FCV-68-308/146.98 0.23 0.16 540.08 2-FCV-68-308 5.3 EFFECT OF SEAT LOAD REDUCTION ON SEAT LEAKAGE FOR METAL SEATED VALVES Unlike the soft-seated swing check valves, a tight sealing of a metal-seated valve requires yielding of one material into the surface waviness and surface roughness of the other to block direct leakage paths. Even, seemingly smooth machined surfaces have surface asperities as illustrated in Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 17 Figure 5-1. When the two surfaces contact each other, the surface asperities initially establish the contact. With an increasing load, the asperities initially deform elastically and then plastically. To ensure a reliable seal, the surface asperities within the contact band need to deform plastically over a reasonable amount of bandwidth. At low pressures, the seat load will not be sufficient to plastically yield the asperities on the contacting surfaces and therefore, the asperities will deform elastically. Due to the fact that these valves were providing a reliable sealing at DPtest pressure, the leak path at this pressure will have a very high flow resistance. Any reduction in the seat load will decrease the flow resistance by increasing the leakage flow area, AI, due to a reduced elastic deformation of the asperities (high spots). For the subject valves, the seat load decreases with decreasing test differential pressure. The amount of seat load decrease is limited to the portion of seat load produced by differential pressure. Table 5-2 shows that the seat load reduction for all the valves are below 1% except for Group 32-3. The seat load is reduced by 18.61% for Group 32-3 valves when the pressure is reduced from the normal test DPtest to Pa (16.5 psig to 9 psig). The reduction in the seat load will reduce the leak path flow resistance by increasing the leakage flow area, AI. A reduction in the flow resistance is equivalent to an increase in the leakage coefficient, CL. Based on Equation 3-3 in Section 3.2 of the main report, the CL can increase by 35% before the measured leakage at pressure Pa would increase from the measured leakage at DPtest.A calculation is performed for Group 32-3 valves to determine the percentage increase in the leakage flow area, AI, when the pressure is reduced from DPtest to Pa (16.5 psig to 9 psig). This calculation is performed based on the approach documented in Section 2.5 of Attachment 3 of this report. All the assumptions documented in Section 2.5.2 of Attachment 3 remains the same for the subject globe valves of Group 32-3 accept the assumption for the seat contact band width, t. The seat contact band width, t, for Group 32-3 valve is assumed to be of 0.005 inches which is a reasonable assumption. A lower seat contact band width will provide a conservative result. A sensitivity analysis showed that lowering the seat contact band width to 0.001 inches does not change the overall conclusion because the percentage increase in the leakage area, AI, remains negligible. Therefore, this assumption does not require a verification. The mean seat diameter of Group 32-3 valves is 2.25 inches (Table 3-1).Table 5-3: Percentage Increase in Leakage Flow Area Calculation for Group 32-3 Valves These Variables are Defined in Attachment 3 of the Main Report Valve Group Rr_DPtest, lb Rr_Pa, lb Fhs_DPtest, lb Fhs_Pa, lb ha_DPtest, in ha_Pa, in Nhs AI 32-3 163.28 132.89 1.76E+07 9.29E-06 7.56E-06 0.025 0.020 0.03 Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 18 Table 5-3 shows that the calculated percentage increase in the leakage flow area, AI, is very small and is not expected to increase the leakage coefficient, CL, by 35% which is a threshold for the measured leakage at the lower pressure, Pa, to increase from the measured leakage at the higher pressure, DPtest.Figure 5-1: Microscopic Flow Path Under Light and Heavy Seating Load [6]5.4 NOTES/ RECOMMENDATIONS The minimum recommended seat contact force value for metal seats is 100 lb/in per Reference 6.Per Table 5-1, the seat contact force values at 16.5 psig (DPtest) and 9.0 psig (Pa) pressures exceed 100 lb/in except for Group 32-3 valves. Although, the calculated seat load for Group 32-3 is below the recommended 100 lb/in at DPtest, the historical measured leakages for Group 32-3 (Unit 1) valves are below the acceptable value [3].

~~5.5 CONCLUSION~~

As can be seen from Table 5-2, the seat load reduction for all the valves except for Group 32-3 are below 1%. The seat load is reduced by 18.61% for Group 32-3 valves when the pressure is reduced from the normal test DPtest to Pa (16.5 psig to 9 psig). The maximum test DP that results in 10%seat load reduction for Group 32-3 is 12.65 psig. Per Section 5.3, to be at risk for increased leakage at a Pa = 9.0 psi, the reduction in the seat load would need to increase the leakage coefficient, CL, by 35%. Based on the simplified but conservative calculation performed in Section 5.3, it is expected that the change in the leakage flow area, AI, will be negligible (see Table 5-3) with the reduction in seat load. Therefore, it is expected that the leakage coefficient, CL, will not increase by 35% which is a threshold for the measured leakage at the lower pressure, Pa, to increase from the measured leakage at the higher pressure, DPtest. Therefore, the seat load reduction of 18.61%for Group 32-3 valves is not expected to increase the leakage at pressure Pa.Based on these results, seat leakage for all AOV Glove valves within the scope of this assessment is not expected to increase if tested at a lower differential pressure of 9.0 psig.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 19 6REFERENCES

1. KEI Document No. 1500C Rev. 15; Kalsi Engineering, Inc. Quality Assurance Manual.
2. TVA Purchase Order 6232543, Rev. Num: 0.
3. TVA Engineering Work Request, EWR20MEC088032, Generate List of U1 Containment Isolation Valves for Kalsi Engineering Pa impact evaluation. 06/09/20.
4. Evaluation Guide for Valve Thrust and Torque Requirements. EPRI, Palo Alto, CA: 2016.

3002008055.

5. Nuclear Maintenance Applications Center, Application Guide for Motor-Operated Valves in Nuclear Power Plants - Revision 2, Volume 1: Gate and Globe Valves, EPRI, Palo Alto, CA, August 2007, 1015396.
6. ISA Handbook of Control Valves, 2nd Edition, J.W. Hutchison, Instrument Society of America, 1979.
7. TVA Calculation:
a. MDQ00099920040092, Rev. 10, Set Point Controls Parameters Review Calculation for Watts Bar Category 2 Air Operated Valves (AOVs), Dated: 06 20.
b. MDQ00299920090344, Rev. 12, Set Point Controls Parameters Review Calculation for Watts Bar Category 2 Air Operated Valves (AOVs), Dated: 09 17.
8. TVA Engineering Work Request, EWR20MEC026076, Work Order # 121532992, Generate U2 Containment Isolation Valve List and Design Inputs for Kalsi Engineering Pa Impact Evaluation, 08/19/20.
9. WBN UFSAR Section 6.2, Containment Systems.
10. Valve Drawing:
a. Lesli Co., Diaphragm Control Valve 2 Class DDOSX 150# S.W.E., Dwg. No.

717543070D, Rev. 901, Dated: 10-27-05.Non-Proprietary Version

Document 3960C, Rev. 0, Attachment 5 Page 20

b. Fisher Controls, Air Operated Control Valve, Type 667-SS-95, Size 1 Body 40 Actuator. Dual Spr., Dwg. No. 54A0223, Rev. 902, Dated: 11-01-88.
c. Fisher Controls, Air Operated Control Valve, Type 667-SS-95, Size 1 Body 40 Actuator. Dual Spr., Dwg. No. 54A0240, Rev. 902, Dated: 11-01-88.
d. Fisher Controls, Air Operated Control Valve, Type 667-SS-95, Size 3/4 Body 40 Actuator. Dual Spr., Dwg. No. 54A0237, Rev. D, Dated: 11-17-80.
e. Copes-Vulcan, Inc., 3/8 Class 1500 ASME Valve Assembly Code Class-2, Dwg.

No. E-171555, TVA Rev. 902, Dated: 9-18-91.

~~11. ANSI/ANS-56.8-1994, American National Standard for Containment System Leakage Testing Requirements.~~

Non-Proprietary Version

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 1 of 14 Appendix A SUPPORTING DOCUMENTS Page No.Title Page 1A Reference 6 2A Reference 10 10A Total Pages 14A Non-Proprietary Version

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 2 of 14 96 VALVE TRIM Application tact the seat joint as the seating load is

~~1) Simpler to apply, because of fewer applied by the actuator.~~

restrictions, such as flow direction, If the _Plu~ strike~ the seat joint slightly air supply, orifice size, pressure drop, cocl~ed, it will rem am co*cked until a higher etc. seatmg load causes it to jump sideways

2) Fewer requirements for positioners. as it slides down into the taper and slams
3) A hydraulic snubber is never re- into full joint contact. This deforms the quired. ~eat, causin~ leakage. In time, the plug
4) Quick change trim. mdenture will extend to form a new off center, nearly 360 ° seat contact. Above SEATING, SEALING AND LEAKAGE a certain plug co*cking angle, the plug The three problems discussed in this will not jump into place regardless of section include: loading; therefore, pre-guiding and align-Seating - The alignment and contact ment of the plug before seating, is neces-of the plug with the seat, in- sary.

cluding joint design and load- The two principal types of mis-alignment ing. are Cori:centric as s~own in Figure 68( a)Sealing - The parameters of metal fin- and Axial as shown m Figure 68(b ).ish, joint width, and metal yielding which lead to tight Alignment sealing. Alignment of the plug on the seat for a Leakage - The amount of leakage that single ported, top and bottom guided may be expected for different ':alve involves concentric alignment of sealing parameters and joint eight components and axial alignment designs. of eleven, fit combinations; plus consider-

~~. Tight sealing is becoming of greater ation of the operating clearances.~~

nnportance to control valve users, now One can readily see the precision ma-that improvements in designs of both chining, required of the control valve valve trim and actuators allow tight shut- manufacturer to maintain alignment of off. One valve may be used for both stop these parts. Each part must have an a~d throttling service at a cost saving. assembly or sliding clearance which allows Diaphragm control, valve actuators with a minute horizontal axial shift and a very 15 to 35 psi air supplies do not develop sligh~ co*ck. The flexibility of the stem is the high seating forces used in stop valve sufficient to allow the plug to move into designs with manual or automatic opera- true seat joint contact with light, initial tion. If they did, they would be bulky seat-contact loading.and would be sluggish in their speed of stem movement for throttling action. A small amount of leakage is to be ex- Seat Joint Design pected, because of the lower seating forces.Manufacturers normally rate their valves Flat joints, normal to the plug axis, for maximum leakage as follows: are used on some low pressure stop valves and soft seated control valves. It is not Double Seated Valves - practical to manufacture them for high

<0.1% Cv maximum I pre~sure service. As discussed later, tighter s~ali_ng occ~rs through a sliding and bur-Single Seated Valves - mshmg action of a tapered joint. Tapered <0.01% Cv maximum! joints turn the fluid gradually and are the best _for high pressure drop and for erosive Development of the springless piston service. The control valve industry uses actuator, air loaded on both sides and joints from 15. to 45°. Smaller angles 0

using a much higher supply pressure would begin fo fo rm sticking tapers and( 100-150 psi), coupled with a positioner larger angles give too much of a poppet led to higher seating forces and tighte; effect, when cracked open at high pres-shut-off. Single seated valves can now be sure differentials, which would cause un-sealed, drop-tight to high pressure drops. stable -rangFtl'frottling. l fi" summary:Some cage balanced valve designs also 45° sea ang es md their .best appli-shut off drop-tight with small diaphragm cations in either normally open or actuators. normally closed valves.

2) 30~ seat angles are a compromise Seating between high seating forces and To make a seal, a plug must first be streamlined flow, at low lifts, for low perfectly aligned and then must fully con- erosion service.

1 se_at ~e~age.flow islaminarratherthanturbuf ent and the C:v formula is not applicable; therefore, this is simply a means of specifying an amount of acceptable leakage.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 3 of 14 SEATING, SEALING AND LEAKAGE 97 4.,SEAT PORT t_PLUG SEAT DIAMETER AT PLUG CONTACT (a)( b)NON CONCENTRIC co*ckED AXIS Figure 68. PLUG MISALIGNMENT WITH SEAT.

3) 15 ° seat angles are best applied to (0.01% Cv maximum); willseal3000 high pressure drop, erosive service. ___:,, P-Si pressure drop on 0.015 in. wiam,
ffio Jomt of 316 SSl, Seat Joint Width 4) 300 lb.fin. - Very high pressure drop Seat joint width *is a balance of adding service; drop-tight (will seal 6000 length to the flow path to increase flow ~ on 0.025 in. width, 20° joint"or resistance vs. reduction of the seating force 440-C SS, hardness 55 Re)*

for a given actuator seating load. A cer- 5) 600 lb. f in. - Extremely high pres-tain minimum width is essential to es- sure service.tablish a tight seal; however, the joint must have sufficient backup strength to The apparent compressive seating stress-support the compressive seating load and es on joints described in items 3) and 4) must be wide enough to prevent inden- above, are 13,000 psi and 35,000 psi tion of the plug. Narrow joints are much respectively, which is well below the yield tighter than wide jomts, provided they point of the given trim materials. These exceed the minimum width requltements. are the stresses normal to the joint. Elasto-L

.Ainerk and plastic yielding is occuring at the "' fiigl1 points of each surface making a tor-Seat Joint Loading S eat joint loading is usually expressed as pounds of force per linear inch of tuous flow path for leakage restriction. er, By contrast, stop-valve seat loading with ( !X.tW mean seat joint circumference. Loading g a rd-faced seats in steam service may be ve.

may vary from 25-600 lb. f inch as given 0,000 psi, appa.rent stress, or £.our ti.mes ~below: a nominal control valve, seating load of /

~~1) 25 lb.fin. - Low pressure drop ser- 100 lb/linear inch for a diaphragm ac-vice; leak-tight shut-offis not required; tuator.~~

metal-to-metal joint. To obtain the best circular seat-joint

~~2) 50 lb.fin. - IModerate pressure drop contact at low stem loading use:~~

service; slight leakage expected ( 0.1%Cv maximum). 1) Cage guided trim with the seat in-

~~3) 100 lb.fin. - High p ressure drop tegral with the cage. Horizontal and C~~

service; nearly "drop-tight servi<;e axial alignment for seating involves 1 100 lbs/li~ear inch vertical seating force is equivalent to a force of 200 lbs/linear inch normal to a 30° joint (100 lb./sin 30°). The compressive seating stress is 200lb/0.015in.x lin.=13,300psi.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 4 of 14 98 VALVE TRIM only two parts in sliding contact. Trim Sealing Alignment of regular trim must be transferred from the seat to the body, Tight sealing requires the yielding of to the bonnet, to the guide, to the one material into the surface "waviness" plug, and back to the seat. and surface roughness of the other, as

2) Cage guided trim with the seat aligned illustrated by Figure 70, to block direct by the cage. leakage paths; thus, making these paths
3) A seat that is integral with body; or long and tortuous. Compressive stresses
4) Extend a thin flexible seat lip above are far below those that would yield the the means of seat retention as shown entire joint; therefore, the contact area is in Figure 29. apparent. Actually, only the peaks of each A radial expandable, seat ring design is surface are in contact and the concentra-illustrated by Figure 69. In this design the tion of force may then exceed the yield seating angle creates a large radial com- and will plastically deform the high spots ponent of force which expands the seat on each valve closure. Additional closures ring against the retaining collar. The require a higher seating load to achieve spring-out action of the non-circular ring the same degree of tightness, until the allows a near-perfeci joint contact giving wear particles are formed and conditions drop-tight shut-off in severe thermal cycle tend to stabilize.

service. This design has been successfully used on pressure drops to 4,400 psi. The Tapered joints provide for a sliding and large flow, entrant passage also makes burnishing action as contact is made and the valve suitable for erosive service. loading occurs. This gives a tighter initial seal than a perpendicular contact and the seal remains tighter with repeated closures.The minimum width of a joint to seal gas to 1 x 10-7 cc/s/linear inch, maxi-mum leakage, is 0.04 inch. Wider:-:-* 0 *-::_---------,with the same surface finish and loading will not seal tighter. This width insures sufficient high point contact to form an adequate flow resistance path as shown by the graph of Figure 71.Extra "super-finishing" of seat joints is unnecessary for tight sealing, because as the joint opens and closes, wear parti-cles ball up on the surface, quickly re-turning it to a rougher finish. Also, some fluid contaminants tend to remain in the joint on closure and indent the surfaces.Figure 69. RAD I All V EXPANDABLE SEAT Excessive lapping generally either reduces RING DESIGN. seat tightness by increas-ing the actual contact area, thus, reducing the unit com-Courtesy Corwflow Corporation pressive seating stress provided by a fixed actuator seating force or it destroys the original surface geometry.ROUGHNESS HEIGHT (J.16 MICRDINCH){lAYOFROUGHHESS CONCENTRIC ABOUT PlUG AHO SEAT HIS)(LA'l'OF ROUGHNESS CONCENTRIC ABOUT PLUG AND SEAT AXIS)(a) Under Light Seating losd {b) Undar Heavy Seating lo~<l fig1me 70. MATING OF SEAT JOINT SURFACES,

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 5 of 14 SEATING, SEALING AND LEAKAGE 99 0 10*5 a::Cot-0 zz Zg-c..,o I - ~ ....10*6

~~~~~ a::~~::c c:,t-c.>~~

c~Z 10"7

~~io::::E-c.:=-J~~
~~~zc~~

_,~~

~~.._u~~~

o..,_, 10-*c:,iiiia:: 0 .03 .06 .09 .12 o=-~_,.., ~WIDTH OF JOINT ROUGHNESS PEAKS BAL LEO PARTICLES OF SOFTER SEATING MATERIAL ARE REDU CED WELDED TO HARDER SURFACE OR FREE IN JO INT. ALSO FLUID Figure 71. MINIMUM JOINT WIDTH CONTAM INANT PARTICLES ARE IN THE JOINT.FOR A TIGHT SEAL.(Leakage rates for 14.7 psi AP helium on a flat Figure 72. JOINT SURFACE AFTER circular joint.) REPEATED CLOSURES.Consolidated graph taken from Reference No. 48.1000,-----------.-------~----------,M 100

u z REGION OF 0 GOOD SEAT ai::u FINISH ii MM z 16 u=

0 ai::10 SUPER FINISH uC WILL NOT HOLD UP

ai::M u0 11 10 90 100 1000 LOG WAVINESS HEIGHT MICROINCHES

~~NORMAL TO LEAKAGE PATH Figure 73. DEGREE OF SEAT JOINT FINISH VS. METHOD OBTAINING FINISH.~~

Chart presented in "Investigation of Leakage and Sealing Parameters", Paul Sauer, ITT Research liistitute, Technical Report AFRPL-TF 153, August, 1965.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 6 of 14 100 VALVE TRIM The following comparison illustrates the and erosive fluids at 1000 to 4000 psi relative seating force to obtain same de- pressure drop. Few stop valves could gree of tightness vs. the seat joint finish- match this performance and remain tight.ing method ( concentric lay of finishing It is customary to expect a slight degree pattern). Refer to Figures 72 and 73. of leakage and, in general control valve service applications, this has presented few problems. Specified permissible leak-Finishing Method Relative Seating Force age should be based on process consider-diamond burnished 1.0 ations (hazards resulting from leakage (after grinding) after emergency shut off, etc.). Tight shut off should be specified only when neces-lathe turned 1.3 sary, because a higher quality more ex-pensive valve is required.ground 2.5 Liquid Leakage is to a great extent affected by surface tension and the parti-lapped (excessive) 2.9 cular fluid wetting the joint surface, be-(90% of apparent area of)(contact actually mated) fore the new fluid attempts to enter. For precise leak testing, the joint and body should be free of all traces of oil, which Trim leakage would preferentially wet the metal joint surfaces. Water leakage rates of an oil-Definitions Relative to Trim leakage: free joint are higher. Oil clinging to the roughness, blocks some leakage paths, Drop tight, to be meaningful, must be causing a higher capillary displacement specified in terms of the maximum allowed pressure to establish flow of the test fluid.number of drops per unit of time ( drops/minute, cc/hour or no visible drop).Bubble tight, to gas, should be speci-fied as the maximum allowed number of Seat leakage Specifications bubbles of a given size per minute (usually 1/8 in. diameter). Single Seated Globe Valves Zero leakage is defined as 1 x 1o-8 A maximim allowable leakage of 0.01% of cc/ s or about 0. 3 cc/year (helium leak rate Rated Cv is often based on the nominal or at standard conditions). Zero leakage is catalog listed actuator size for the given valve often specified in critical service and re- tested with 50psi of air across the seat joint.quires very careful joint design, material selection, finish control, and sufficient seat-ing force on narrow joints. It is practical Double Seated Globe Valves only for small valves, at extra cost, and may last for only a few closing cycles. These valves are specified to have a max-Stop valve maximum leakage rates are imum leakage rate of0.5% to 0.1% ofrated Cv given in the Valve Manufacturers Stan- depending upon quality purchased. They are dardization Society, SP-61 as: tested with 50 psi of air across the seat joint.

~~1) Water tests at a P = Body CWP rat- See ISA standard 39.1.~~

ing (10 cc/hour/inch of valve pipe size or about 3/drops/minute). Extra Tight Shut Off for

~~2) Air tests at l:..P=80 psi (0.1 SCFM/ Single Seated Globe Valves inch of valve pipe size).~~

Leakage Specifications, to be meaningful A water test is often conducted at the differ-and allow comparison, should includetest ential pressure rating assigned to the valve fluid, temperature, pressure, pressure by the manufacturer and. which is based upon drop, seating force and duration of test. actuator size, air loading, spring force and direction of leakage across the plug (either Control Valve leakage over or under the plug). Maximum leakage may be specified as 0.0005 cubic centimeters Properly designed control valves can of water per minute per inch of valve seat achieve stop valve tightness and maintain orifice diameter (not pipe size of valve end) it throughout a long service life before per psi pressure drop. Example: A valve hav-trim replacement; particularly with cage ing a 4 inch seat orifice and tested to 2000 psi guided, balanced trim having elastomer plug-to-cage seals. The control valve, how- differential pressure would have a maximum ever, is expected to throttle and often shuts water leakage rate of 4cdminute. Leakage off much more frequentlythanstopvalves. may be checked with a gas instead of a liquid.For example, some dump valves may have The maximum allowable rate is often from 4000 to 7000 opening and closing specified at 6 x 10- 1 cubic centimeters per sec-cycles per day, handling high pressure ond per inch of seat diameter.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 7 of 14 SEATING, SEALING AND LEAKAGE 101 Soft Seats in Globe Valves These may be leak tested in accordance with SAMAStandard#PMC23.2c.Underthis h standard valve actuators are sized to provide a minim um stem force of 150 pounds per inch of seat diameter over and above the plug un-balance force created by the maximum rated differential pressure. The test pressure using air is the maximum rated differential pres- for gases the leakage formula becomes sure or 300psi, whichever is the least, but not to exceed the maximum operating pressure at 1r h 3 2 2](ro+.ri) (P1 -.P2 )ambient temperature.Elastomer Lined Butterfly Valves[Qo = - - - - - - - - .12µ (r0 - q) Po XThese are often tested with 50psi of air as a minimum, or maximum differential operat-ing pressure across the disc. With the down-stream side of the disc covered with water, the maximum allowable leakage rate may be specified at one bubble of air in ten seconds, Ao= molecular mean free path at stan-per inch of vane diameter. dard conditions.E = correction factor, 0. 9 for a single gas and 0.66 for a mixture.Elastomer Sealed Ball and Plug Valves They are usually bubble tight to their rated differential pressure. Metal seated valves The problem is that the leakage test have relatively high leak rates compared to gives no indication of whether the leak globe style valves. One exception is a rotary path is one large scratch or the sum of leakage through millions of tiny tortor-cam type plug valve with leakage rates com- ous paths. Refinishing the joint may eli-parable with globe valves. minate the first cause, but the condition may already be at the practical limit of Theoretical leakage Formula seal tightness for the latter.The equation relating the fluid proper- Type Of Gas Flow Through Seat Joint ties, flow path geometry, and flow rate is: leakage Pattern 2 2 -3 The following is a summary of charac-71" (p -p )h teristics for various leak conditions.1 2 (See 1 ) Minimum Restricting Dimension in Plug Position Type of Flow Leakage Path where: cracked open nozzle flow >0.005 in.uniform channel clearance 2 seating load turbulent 0.0005 to pressure at standard conditions build up zone channel flow 0.005 in.exit fluid pressure inlet fluid pressure valve seated and laminar flow 0.0001 to leakage begins 0.0005 in.volume rate of flow at standard conditions joint surface transition flow 0.000001 to outside and inside radii of waviness, then (molecular and 0.0001 in.joint sealing area roughness provides laminar)

~~µ absolute viscosity leak paths as deformation begins When the terms are rearranged, the elastic and plastic molecular flow <0.000001 in.~~

uniform channel clearance is: yielding of joint has closed large paths 1Taken from Reference 50, the formula applies to a flat seat joint.2 This ii; the theoretical separation of two truly plane surfaces to give an equivalent rate of leakage caused by channels, imperfections, etc.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 8 of 14 102 VALVE TJJ,IM (d) Lightly restrained O-rings for low pressure drop, installed on double V-port plug.(a) Nylon seal retained in plug.Courtesy Fisher Controls Company (b) TFE seal in plug of cage guided, balanced valve.(e) O-ring in dove-tail groove for higher pressure drop.Courtesy Fisher Controls Company Vent holes relieve pressure under O-ring retaining it in the groove as the seat joint cracks open. Holes are small to prevent seal extrusion.METAL* TO* METAL UCK*UP SEAL ELASTOMER SEAL RING PRIMARY SEAL (I) Soft seat design used in boiler leedwater pump recirculation anticavitation low noise valve.Handles fluids to 475° Fend 6000 psig inlet pressure.(c) Elastomer seal in seat ring of split body valve. Courtesy of Masoneilan International, Inc.Figure 74. SOFT SEAT RETENTION.

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 9 of 14 SEAT ATTACHMENT 103 When sealing to the point of achieving A secondary advantage of soft sealing molecular flow, the lighter, smaller and is that the seal, once compressed, is free higher velocity molecules, such as helium to re-expand and to follow seat distortion and hydrogen, will have the highest leak- as the pressure drop loading increases.age rates. To do this, the sealing material should have a rapid recovery rate upon removal of the load. This action will occur only at Determination of Seat leakage Rate resilient temperatures.Determining the Seat Leakage Rate is not Hard materials such as nylon, in thin as simple as it may appear. Fluids are sections carefully restrained, can handle subject to thermal expansion and contrac- pressure drops of several thousand psi; tion during tests and air may go in or whereas, TFE materials are readily out of solution, causing volume changes abraded.in the downstream measuring system. Soft seals are useful for sealing where For accurate testing, maintain the valves contaminants or solid material are trapped and the fluids at an equal temperature in the closed joint. Material as hard as and air dissolved in water at equilibrium 60D in a raised-seal headform is capable conditions. of sealing 0.01 in. particles, bubble-tight, Helium leak detection requires a clean to 1500 psi pressure drop.background with large amounts of fresh The softer the seal, the better its abrasive air for maximum sensitivity. resistance, up to the point where it is The following is a summary of the damaged by pressure drop forces.sensitivity of various test methods, given Large volume, high pressure blowdown in cc/s at atmospheric conditions. systems are necessary to adequately test elastomer seals and retention means for air bubbles in water 1 x 10-3 to 1 x I0-4 cc/s pressure drop strength. Elastomers, under (also air and high loading, tend to act as a fluid and soap bubbles) extrusion may occur unless the load is limited or unless a metal-to-metal stop is thermal detectors 1 x 10-4 cc/s used. Some joint designs allow soft sealing halogen detectors 1 x 10-5 cc/s first, followed by a metal-to-metal closure as a secondary seal in case of soft seal mass spectrometer rupture.using "sniffer" 1 x 10-6 to 1 x 10-8 cc/s The material properties to beconsidered (helium leak pick up probe) in the selection of a soft seat are:

1) Fluid compatibility including, swell-ing, loss of hardness, permeability, degradation;
2) Hardness; Soft Seating 3) Permanent set;
4) Rate of recovery upon removal of Resilient composition sealing materials load; are used to obtain bubbletightsealingwith 5) Tensile and compressive strength; a small actuator force. Compressive seat- 6) Distortion before rupture; ing stresses are such, that the material is elastically deformed into the surface 7) Modulus of elasticity.

roughness of the mating metal part, to Rarely do the physical properties given block all leak paths. The permeability for sealing materials relate to the actual of the material, to the fluid, is the source conditions of loading and strain in valve of a very minor leakage. seals. There is no substitute for thermal Materials which are too soft, or that tend and blow-down testing as the means to to cold-flow (creep) under load, may be prove material, seal configuration and stiffened with fillers such as glass. When joint design.used in thin sections, and adequately re-tained, the cold-flow or permanent-set SEAT ATTACHMENT problems may be eliminated. AND SEALING TO BODY Seals must be carefully restrained against rupture and blow-out by differen- Seat attachment and sealing to the body tial pressure. Several designs of soft-seat is a major consideration of valve sealing, retention are illustrated by Figure 74. equal in importance to joint seal, bonnet The bonding of seats to metal parts is an seal, and stem seal. Lack of seat-to-body aid, but not a total solution, because sealing gives a continuous leak, often bonds are subject to thermal shock crack- blamed on the seat-to-plug joint. In high ing and to degradation. Sufficient pressure pressure and/or steam service, leakage drop will rupture the bonding material. behind the seal will actually erode through

Appendix A Document 3960C, Rev. 0, Attachment 5 Page 10 of 14

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Enclosure 5 Kalsi Engineering Affidavit CNL-20-006

APPLICATION FOR ORDER APPROVING LICENSE TRANSFER AND CONFORMING LICENSE AMENDMENT 10 CFR 2.390 AFFIDAVIT Affidavit of Manmohan S. Kalsi I, Manmohan S. Kalsi, President and Officer ofKalsi Engineering, Inc., do hereby affirm and state:

1. I am authorized to execute this affidavit on behalf of Kalsi Engineering, Inc. (KEI).
2. KEI requests that Report 3960C and Attachments, which are labeled "Proprietaiy and Confidential" per 10 CFR 2.390 (b)(l)(i)(A), be withheld from public disclosure under the provisions of 10 CFR 2.390(a)(4). Another version of this report marked "Non-Proprietaiy" has been provided for public records.
3. Repo1i 3960C and Attachments contains confidential info1mation, the disclosure of which would adversely affect KEI.
4. This information has been held in confidence by KEI. To the extent that KEI has shared this information with others, it has done so on a confidential basis.
5. KEI customarily holds such information in confidence since the information is not available to the public to the best of our knowledge and belief.
6. Public disclosure of this info1mation would cause substantial harm to KEI's business opportunities because such information has significant commercial value to KEI and its disclosure could adversely affect these opportunities by making it available to our competitors.

_on_~-----'--A-*_KJu_,_ ~ 15/2. o'l.0

~~Manmohan S. Kalsi President, Kalsi Engineering, Inc.~~
~t':f.!~?l~ NANCY A. RICHEY ;f(:,ti.:~'s Notary Public, State of Texas \.~. *..~~ Comm ._Expires 03-26-2024 ..,. ..;_(f'" **~:$' ~,,,,ftr,,,,," Notary ID 124872147 Subscribed and sworn before me, A Notaiy Public This ['5fh day of ~ ~ Y , 2020.}}

Hinkley Point B nuclear power station
Reactor type	GCR - AGR
Reactor supplier	General Electric Company, Whessoe, and Strachan & Henshaw (Graphite supplied by Anglo Great Lakes Corporation Ltd)
Cooling source	Bridgwater Bay
Thermal capacity	2 x 1494 MW_t