CNL-14-176, Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03)

Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03)

ML14339A539
Person / Time
Site:	Sequoyah
Issue date:	12/02/2014
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-14-176
Download: ML14339A539 (213)

Text

1101 Market Street, Chattanooga, Tennessee 37402 CNL-14-176 December 2, 2014 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject:

Application to Revise Technical Specification 6.8.4.h, "Containment Leakage Rate Testing Program," (SQN-TS-14-03)

References:

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, ADAMS Accession No. ML12221A202.

2. Electrical Power Research Institute (EPRI) report, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" (EPRI Product No. 1018243), dated August 1994.

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 2, May 2011, ADAMS Accession No. ML13109A112.

4. TVA Letter to NRC, "Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10)," dated November 22, 2013 (ADAMS Accession No. ML13329A717)

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2.

U. S. Nuclear Regulatory Commission Page 2 December 2, 2014 The proposed changes would revise the SQN, Units 1 and 2, Technical Specifications (TS) 6.8.4.h, "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 (ILRT) interval from 10 years to 15 years and the Type C local leakage rate testing intervals from 60 months to 75 months.

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" (EPRI Report 1018243)

(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).

By letter dated November 22, 2013 (Reference 4), TVA submitted a proposed amendment request converting the SQN, Units 1 and 2, TS to the improved Standard TS of NUREG-1431, "Standard Technical Specifications - Westinghouse Plants." The changes proposed in this amendment request are separate from the changes proposed in Reference 4. The changes proposed in this amendment request are based on the current licensing basis TS. Upon approval of the TS conversion requested in Reference 4, revised improved TS pages incorporating the changes proposed in this amendment request will be provided to the U. S. Nuclear Regulatory Commission (NRC) Licensing Project Manager. to this letter provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachment 1 to the enclosure provides the existing TS pages marked-up to show the proposed changes. Attachment 2 to the enclosure provides the retyped TS pages incorporating the proposed changes. Enclosure 2 to this letter is a risk impact assessment for permanently extending the containment Type A test interval. Enclosure 3 to this letter provides the resolutions of the Probabilistic Risk Assessment Peer Review Team Facts and Observations.

TVA requests approval of the proposed license amendment by November 1, 2015, in order to incorporate these changes into the SQN schedule. The license amendment will be implemented within 60 days of NRC approval. This would allow deferral of the next ILRT Type A tests, currently scheduled for the Unit 2 Cycle 20 refueling outage in the fall of 2015 and the Unit 1 Cycle 21 refueling outage in the fall of 2016. The last Unit 1 ILRT was completed on October 27, 2007, and the last Unit 2 ILRT was completed on December 20, 2006.

The proposed changes allow the ILRT to be performed within 15 years from the last test on each unit. This application represents a cost-beneficial licensing change. The

U. S. Nuclear Regulatory Commission Page 3 December 2, 2014 ILRT imposes a significant expense on the station while the safety benefit of performing it within 10 years versus 15 years is minimal.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

The SQN Plant Operations Review Committee and the TVA Nuclear Safety Review Board have reviewed this proposed change and determined that operation of SQN in accordance with the proposed change will not endanger the health and safety of the public.

Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

There are no regulatory commitments associated with this submittal. Please address any questions regarding this request to Mr. Edward D. Schrull at (423) 751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1st day of December 2014.

J. he a

• resident, Nuclear Licensing

Enclosures:

1. Evaluation of Proposed Change

2. Risk Impact Assessment

3. Resolutions to Probabilistic Risk Assessment Peer Review Team Facts and Observations Enclosure cc (Enclosures):

NRC Regional Administrator- Region II NRC Resident Inspector- Sequoyah Nuclear Plant NRC Project Manager- Sequoyah Nuclear Plant Director, Division of Radiological Health- Tennessee State Department of Environment and Conservation

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 EVALUATION OF PROPOSED CHANGE

Subject:

Application to Revise Technical Specification 6.8.4.h, "Containment Leakage Rate Testing Program," (SQN-TS-14-03)

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. BACKGROUND

4. TECHNICAL EVALUATION 4.1 Description of Containment 4.2 Integrated Leak Rate Test History 4.3 Type B and Type C Testing Programs 4.4 Supplemental Inspection Requirements 4.4.1 IWE Examinations 4.4.2 IWL Examinations 4.5 Deficiencies Identified 4.6 Plant-Specific Confirmatory Analysis 4.6.1 Methodology 4.6.2 PRA Technical Adequacy 4.6.3 Conclusion of Plant-Specific Risk Assessment Results

5. REGULATORY EVALUATION 5.1 Applicable Regulatory Requirements/Criteria 5.2 Precedent 5.3 No Significant Hazards Consideration 5.4 Conclusions

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES ATTACHMENTS

1. Proposed TS (Markups) for SQN, Units 1 and 2

2. Proposed TS Changes (Final Typed) for SQN, Units 1 and 2

3. Proposed improved TS Page Mark-Ups for SQN, Units 1 and 2 Page E1-1 of 27

1.0

SUMMARY

DESCRIPTION This evaluation supports a request to revise the current licensing basis of Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN),

Units 1 and 2, by revising the SQN, Units 1 and 2, Technical Specifications (TS) 6.8.4.h, "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 (ILRT) interval from 10 years to 15 years and the Type C local leakage rate testing (LLRT) intervals from 60 months to 75 months.

2.0 DETAILED DESCRIPTION This proposed change in the current licensing basis is a permanent change of the Type A test interval from 10 years to 15 years and of the Type C tests from 60 months to 75 months. The approved change of the ILRT and LLRT would be incorporated into TS 6.8.4.h by modifying the existing first paragraph of TS 6.8.4.h to reflect the new revision of NEI 94-01.

The first paragraph of SQN, Units 1 and 2, TS 6.8.4.h states:

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. Visual examination and testing, including test intervals and extensions, shall be in accordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing instructions and the following:

The first paragraph of SQN, Units 1 and 2, TS 6.8.4.h is revised to state:

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. Visual examination and testing, including test intervals and extensions, shall be in accordance with NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated 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, with exceptions provided in the site implementing instructions and the following:

Attachment 1 to this enclosure provides the existing TS pages marked-up to show the proposed changes. Attachment 2 to this enclosure provides the retyped TS pages incorporating the proposed changes.

By letter dated November 22, 2013 (Reference 2), TVA submitted a proposed amendment request converting the SQN, Units 1 and 2, TS to the improved Standard TS of NUREG-1431, "Standard Technical Specifications - Westinghouse Plants." The changes proposed in this amendment request are separate from the changes proposed Page E1-2 of 27

in Reference 2. The changes proposed in this amendment request are based on the current licensing basis TS. However, Attachment 3 to this enclosure provides the proposed SQN improved TS pages marked-up to show the proposed changes as an aid in the review. Upon approval of the TS conversion requested in Reference 2, revised improved TS pages incorporating the changes proposed in this amendment request will be provided to the U. S. Nuclear Regulatory Commission (NRC) Licensing Project Manager.

3.0 BACKGROUND

The testing requirements of 10 CFR 50, Appendix J (Reference 3), 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 6.8.4.h.

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 (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 ILRT imposes significant expense on the station while the safety benefit of performing it within 10 years, versus 15 years, is minimal. The benefits of 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 no reduction in 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 4) was issued. RG 1.163 endorsed NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 94-01, Appendix J" (Reference 5), with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT Page E1-3 of 27

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 6) and Electric Power Research Institute (EPRI) Topical Report (TR)-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals" (Reference 7), both of which illustrated that the risk increase associated with extending the ILRT surveillance interval was very small.

NEI 94-01, Revision 2 (Reference 8), 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. It 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, it 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 9), documents the NRC's evaluation and acceptance of NEI 94-01, 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 10).

EPRI Report 1009325, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," Revision 2 (Reference 11), provides a risk impact assessment for optimized ILRT intervals of up to 15 years, utilizing current industry performance data and risk-informed guidance, primarily Revision 1 of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Bases" (Reference 12). 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 13).

NEI 94-01, Revision 3 (Reference 14), 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.

This method uses industry performance data, plant-specific performance data, and risk Page E1-4 of 27

insights in determining the appropriate testing frequency. 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 15), 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).

4.0 TECHNICAL EVALUATION

As required by 10 CFR 50.54(o), the SQN, Units 1 and 2, containments are subject to the requirements set forth in 10 CFR 50, Appendix J. Option B of Appendix J requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Currently, the SQN 10 CFR 50 Appendix J Testing Plan is based on RG 1.163, which endorses NEI 94-01, Revision 0. This LAR proposes to revise the SQN 10 CFR 50, Appendix J Testing Plan by implementing the guidance in NEI 94-01, Revision 3-A.

In the SE issued by the NRC dated June 8, 2012, the NRC concluded that NEI 94-01, Revision 3, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS with regard to containment leakage rate testing for the optional performance-based requirements of Option B of 10 CFR 50, Appendix J.

The following addresses each of the limitations and conditions of the 2008 and 2012 SEs.

Limitation / Condition TVA Response (from Section 4.1 of SE dated June 25,2008)

1. For calculating the Type A leakage rate, the Following NRC approval of this LAR, SQN, Units 1 licensee should use the definition in the NEI and 2, will use the definition in Section 5.0 of NEI 94-TR 94-01, Revision 2, in lieu of that in 01, Revision 3-A, for calculating the Type A leakage ANSI/ANS-56.8-2002. rate when future SQN, Units 1 and 2, Type A tests are performed. The definition in Revision 2-A and 3-A is identical.

2. The licensee submits a schedule of containment The schedule of containment inspections is provided inspections to be performed prior to and between in Section 4.4 below.

Type A tests.

3. The licensee addresses the areas of the General visual examination of accessible interior and containment structure, potentially subjected to exterior surfaces of the containment system for degradation. structural problems is conducted in accordance with the SQN, Units 1 and 2, IWE/IWL Containment Inservice Inspection Plans that implement the requirements of the ASME,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a.

4. The licensee addresses any tests and inspections SQN, Units 1 and 2, have already replaced the performed following major modifications to the steam generators. The design change process containment structure, as applicable. addressed the testing requirements. Similarly, any future containment modifications will be addressed by the site design change process.

5. The normal Type A test interval should be less TVA acknowledges and accepts this NRC staff than 15 years. If a licensee has to utilize the position, as communicated to the nuclear industry in provision of Section 9.1 of NEI TR 94-01, Regulatory Issue Summary 2008-27 dated Revision 2, related to extending the ILRT interval December 8, 2008 (Reference 16).

Page E1-5 of 27

beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.

6. For plants licensed under 10 CFR Part 52, Not applicable. SQN Units 1 and 2 are not licensed applications 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 TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.

Limitation / Condition TVA Response (from Section 4.0 of SE dated July 2012)

1. The staff is allowing the extended interval for Following NRC approval of this LAR, SQN, Units 1 Type C LLRTs be increased to 75 months with the and 2, will follow the guidance of NEI 94-01, requirement that a licensees post-outage report Revision 3-A to assess and monitor margin between include the margin between the Type B and the Type B and Type C leakage rate summation and Type C leakage rate summation and its regulatory the regulatory limit. This will include corrective limit. In addition, a corrective action plan shall be actions to restore margin to an acceptable level.

developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

2. When routinely scheduling any LLRT valve Following NRC approval of this LAR, SQN, Units 1 interval beyond 60-months and up to 75-months, and 2, will estimate the amount of understatement in the primary containment leakage rate testing the Types B and C total and include determination of program trending or monitoring must include an the acceptability in a post-outage report, consistent estimate of the amount of understatement in the with the guidance of Section 11.3.2 of NEI 94-01, Type B & C total, and must be included in a Revision 3-A.

licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

To comply with the requirement of 10 CFR 50, Appendix J, Option B, SQN, Units 1 and 2, TS 6.8.4.h currently references RG 1.163. RG 1.163 states that NEI 94-01, Revision 0, provides methods acceptable to the NRC for complying with Option B of 10 CFR 50, Appendix J, with four exceptions described therein. The current SQN, Units 1 and 2, TS 6.8.4.h requires visual examination and testing, including test intervals and extensions, to be in accordance with RG 1.163 with exceptions provided in the site implementing instructions and the following:

BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacity is adequate to maintain Page E1-6 of 27

system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

4.1 Description of Containment The Containment consists of a steel Containment Vessel (SCV) and a separate Reactor Shield Building enclosing the SCV and annulus. The SCV is a freestanding, welded steel structure with a vertical cylinder, hemispherical dome, and a flat circular base that provides Primary Containment. The Reactor Shield Building is a reinforced concrete structure similar in shape to the SCV that protects the SCV from external events.

The SCV for SQN, Units 1 and 2, is a low-leakage, free-standing 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 113 feet 8-5/8 inches in height from the liner on the base to the spring line of the 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 1/2 inch thick at the spring line and the dome varies from 7/16 inch thickness at the spring line to 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 20-foot centers during erection to ensure stability and alignment of the shell. Vertical stiffeners are spaced at four degree arcs.

During the SQN, Units 1 and 2, steam generator replacements, two construction openings were cut into the domes of each units SCV. These construction openings were restored by reinstalling the removed steel sections and rewelding them to the remaining structure using full penetration welds. The integrity of the restored vessel was verified by NDE and leak testing of the welds.

An equipment hatch with an inside diameter of 20 feet has been provided to enable passage of large equipment and components into the containment during plant shutdown. Two personnel access locks were provided for each SCV. Each personnel lock is a welded steel assembly with a door at each end equipped with a double compressible seal to ensure leak tightness of the lock.

4.2 Integrated Leak Rate Test History Previous ILRT testing confirmed that SQN, Units 1 and 2, containment structures' leakage are acceptable, with considerable margin, with respect to the TS acceptance criteria of 0.25 percent (%) of containment air weight at the design basis LOCA pressure (La). Because the last three SQN, Units 1 and 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.

SQN, Unit 1 Test Date As Found Leakage Acceptance Limit*

5/5/90 Mass Point Upper Confidence Limit (UCL) 0.38 of La 1.0 La leakage with penalties Page E1-7 of 27

12/19/93 Mass Point UCL leakage with penalties 0.79 of La 1.0 La 10/27/07 Mass Point UCL leakage with penalties 0.46 of La 1.0 La

• The total allowable "as-left" leakage is 0.75 La, (La, 0.25% of primary containment air by weight per day, is the leakage assumed in design basis accident radiological analyses) with 0.6 La, the maximum leakage from Type B and Type C components.

SQN, Unit 2 Test Date As Found Leakage Acceptance Limit*

3/19/89 Mass Point UCL leakage with penalties 0.32 of La 1.0 La 4/28/92 Mass Point UCL leakage with penalties 1.31 of La** 1.0 La 12/20/06 Mass Point UCL leakage with penalties 0.36 of La 1.0 La

• The total allowable "as-left" leakage is 0.75 La, (La, 0.25% of primary containment air by weight per day, is the leakage assumed in design basis accident radiological analyses) with 0.6 La, the maximum leakage from Type B and Type C components.

• • A single penetration (X-47A) in the ice condenser glycol system was responsible for this failure. Corrective action was implemented to reduce the chance of a recurrence.

Type B and Type C containment penetrations tests (e.g., electrical penetrations, airlocks, hatches, bellows, flanges, H2 Analyzers, and valves) are being performed in accordance with Option B of 10 CFR 50, Appendix J. The current total penetration leakage on a minimum path basis is approximately 31%, for Unit 1, and 16%, for Unit 2, of the leakage allowed for containment integrity (i.e., 0.6 La).

There are no known modifications that will require a Type A test to be performed prior to U1R25 (Fall 2022) and U2R24 (Fall 2021), when the next Type A tests will be performed in accordance with this proposed change. TVA plans to modify the SQN, Units 1 and 2, containment liners to attach a track near the equipment hatch (X-1) on the refuel floor during U1R20 (Spring 2015) and U2R20 (Fall 2015), respectively. This modification is minor and the Type A post-maintenance test will be deferred until the next regularly scheduled Type A test as allowed by NEI 94-01, Revision 0, Section 9.2.4. The associated post-maintenance tests are tracked in the SQN Corrective Action Program and in the surveillance program for completion. TVA is also considering modifications to replace several electrical penetrations on SQN, Units 1 and 2. If replaced, the associated welds on the containment liner will be locally tested and not require a Type A test. 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 0, Section 9.2.4 or NEI 94-01, Revision 3-A, Section 9.2.4, as applicable.

There have been no pressure or temperature excursions in either SQN Unit 1 or Unit 2 containments which could have adversely affected containment integrity since the performance of the last Type A tests. There is no other anticipated addition or removal of plant hardware within either Unit 1 or Unit 2 containments which could affect leak-tightness.

4.3 Type B and Type C Testing Programs The SQN, Units 1 and 2, Appendix J, Type B and Type C leakage rate test program requires testing of electrical penetrations, airlocks, hatches, bellows, flanges, and valves within the scope of the program as required by 10 CFR 50, Appendix J, Option B and Page E1-8 of 27

TS 6.8.4. The Type B and Type C testing program consists of LLRT of penetrations with a resilient seal, hatches, bellows, flanges, and containment isolation valves that serve as a barrier to the release of the post-accident containment atmosphere.

A review of the most recent Type B and Type C test results and a comparison with the allowable leakage rate was performed. The combined Type B and Type C leakage has remained below La (i.e., approximately 225 standard cubic feet per hour (SCFH)) for SQN, Units 1 and 2. The maximum and minimum pathway leak rate summary totals for the last three refueling outages are shown below.

SQN, Unit 1 U1R17 As-Found Minimum Pathway Leakage 13.5 SCFH Fall 2010 As-Left Maximum Pathway Leakage 21.6 SCFH U1R18 As-Found Minimum Pathway Leakage 16.9 SCFH Spring 2012 As-Left Maximum Pathway Leakage 28.5 SCFH U1R19 As-Found Minimum Pathway Leakage 25.1 SCFH Fall 2013 As-Left Maximum Pathway Leakage 78.73 SCFH SQN, Unit 2 U2R17 As-Found Minimum Pathway Leakage 15.4 SCFH Spring 2011 As-Left Maximum Pathway Leakage 31.8 SCFH U2R18 As-Found Minimum Pathway Leakage 18.4 SCFH Fall 2012 As-Left Maximum Pathway Leakage 45.8 SCFH U2R19 As-Found Minimum Pathway Leakage 44.4 SCFH Spring 2014 As-Left Maximum Pathway Leakage 43.6 SCFH Each unit has two airlock doors, 40 individual bellows tests, 47 electrical penetrations, and 14 resilient seal and hatch type penetrations that are LLRT (Type B). Unit 1 has 78 valve penetrations and Unit 2 has 86 valve penetrations that are LLRT (Type C).

Currently, approximately 5% of the components (Units 1 and 2 combined) are tested at the 30-month nominal interval due to performance. However, neither unit's overall Type B and Type C leakage has approached the La leakage limit.

As discussed in NUREG-1493, "Performance-Based Containment Leak-Test Program,"

Type B and Type C tests can identify the vast majority (greater than 95%) of all potential containment leakage paths. This amendment request adopts the guidance in NEI 94-01, Revision 3-A, in place of NEI 94-01, Revision 0, for the Type C test interval (up to 75 months), but otherwise does not affect the scope or performance of Type B or Type C tests. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

4.4 Supplemental Inspection Requirements Prior to initiating a Type A test, a general visual examination of accessible interior and exterior surfaces of the containment system is performed to identify any potential structural problems that could affect either the containment structure leakage integrity or Page E1-9 of 27

the performance of the Type A test. This inspection is typically conducted in accordance with the SQN, Units 1 and 2, Containment Inservice Inspection (ISI) Program, that implements the requirements of ASME, Section Xl, Subsection IWE/IWL. The Code of Record for the examination of Class MC and Class CC components and related requirements for SQN, Units 1 and 2, is the 2001 Edition including 2003 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1 in accordance with 10 CFR 50.55a(g)(4)(ii) and the additional requirements specified in 10 CFR 50.55a(b)(2)(ix)(A), (B), and (F) through (I).

The examinations performed in accordance with the IWE/IWL program satisfy the general visual examination requirements specified in 10 CFR 50, Appendix J, Option B.

Identification and evaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(A). Examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation are performed in accordance with the requirements of 10 CFR 50.55a(b)(ix)(G) and 10 CFR 50.55a(b)(ix)(H). Each ten-year IWE interval is divided into three approximately equal-duration inspection periods. A minimum of one inspection during each inspection period of the IWE interval is required by the IWE/IWL program.

Subsection IWE ensures that at least three general visual examinations of metallic components would be conducted before the next Type A test if the Type A test interval is extended to 15 years. This meets the requirements of Section 9.2.3.2 of NEI 94-01, Revision 3-A and Condition 2 in Section 4.1 of the NRC SE for NEI 94-01, Revision 2.

Subsection IWL provides the examination and exemption requirements for Class CC components. The Class CC equivalent components at SQN, i.e., the base slab and metal liner, are exempt from examination based on the exemptions of IWL-1220(b) and IWE-1220(b).

TVA performs a visual inspection of the accessible interior and exterior of the SQN, Units 1 and 2, Containment Buildings prior to each Type A test. This examination is performed in sufficient detail to identify any evidence of deterioration which may affect the reactor building's structural integrity or leak tightness. The examination is conducted in accordance with approved plant procedures to satisfy the requirements of the 10 CFR 50, Appendix J Testing Program. The activity is coordinated with the IWE/IWL examinations to the extent possible.

Together these examinations ensure that at least three general visual examinations of the accessible containment surfaces (exterior and interior) and one visual examination immediately prior to a Type A test would be conducted before the next Type A test if the Type A test interval is extended to 15 years, thereby meeting the requirements of Section 9.2.3.2 of NEI 94-01, Revision 3-A, as well as Condition 2 in Section 4.1 of the NRC SE for NEI 94-01, Revision 2.

The tables below provide dates of completed and scheduled ILRTs, completed containment surface examinations, along with an approximate schedule for future containment surface examinations, assuming the Type A test frequency is extended to 15 years.

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Unit 1 Calendar Type A Test General Visual Examination of General Visual Examination of Year (ILRT) Accessible Exterior Surface Accessible Interior (Liner) Surface 2006 5/1/2006 2007 10/27/2007 10/11/2007 10/31/2007 2008 2009 4/13/2009 2010 10/23/2010 2011 2012 2/27/2012 2013 10/20/2013 10/27/2013 2014 2015 4/2015 2016 11/2016 2017 2018 4/2018 2019 11/2019 2020 2021 4/2021 2022 11/2022 11/2022 Unit 2 Calendar Type A Test General Visual Examination of General Visual Examination of Year (ILRT) Accessible Exterior Surface Accessible Interior (Liner) Surface 2005 2006 12/20/2006 12/7/2006 12/9/2006 2007 2008 5/22/2008 2009 11/13/2009 2010 2011 6/6/2011 6/4/2011 2012 12/28/12 (PSI, SG replacement) 2013 2014 6/1/2014 2015 12/2015 2016 2017 5/2017 2018 11/2018 2019 2020 4/2020 2021 11/2021 11/2021 4.4.1 IWE Examinations In accordance with the Containment Inservice Testing Program, station personnel perform an IWE - General Visual examination on the accessible surface areas associated with the containment liner. A review was conducted for SQN, Units 1 and 2, per IWE-1241, Examination Surface Areas (1992 Edition with 1992 Addenda of ASME Section Xl) for the initial 10-year Category E-C examination requirements. Four areas on each unit were deemed susceptible to accelerated degradation and aging. Those areas were the inboard and outboard chilled water penetrations (i.e., X-64, X-65, X-66, and X-67). Augmented examinations were completed per the requirements of Category E-C and the components were removed from the augmented examination Page E1-11 of 27

requirements per IWE-2420(c). This information is documented in the first 10-year Containment ISI Plan for SQN, Units 1 and 2.

The 1999 and 2000 examination results, for SQN Unit 2 and SQN Unit 1, respectively, showed pitting and corrosion that required further evaluation. Moisture had been absorbed and held against the nozzle reinforcement by black foam insulation which led to corrosion. Ultrasonic Testing (UT) thickness measurements were obtained and all measurements were greater than minimum wall thickness. There were no visible signs of active corrosion. The areas were cleaned and repainted. Subsequent examinations in 2003, 2006, and 2009 showed the wall thickness to be acceptable and greater than the minimum specified by design calculations.

The examinations performed during the first ten-year interval are characterized by the following examples:

In 2000, a visual examination of the SQN, Unit 1 SCV exterior surface identified minor rusting and discoloration. There were no detrimental flaws or significant degradation noted during the examination. The areas did not affect the structural integrity of leak tightness of the SCV. The areas were cleaned and recoated in accordance with site procedures. A VT-3 preservice examination was performed following reapplication of coatings to satisfy the requirements of IWE-2200(g).

In 2003, a visual examination of the SQN, Unit 2 SCV interior surface identified arc strikes on the SCV. The arc strikes were removed by blending. A visual examination was conducted after their removal and no indications were found. In addition, the surrounding areas were checked for wall thickness loss and the remaining wall thickness was acceptable. The areas were recoated in accordance with site procedures. The areas did not affect the structural integrity or leak tightness of the SCV.

In 2005, a visual examination of the SQN, Unit 2 SCV exterior surface identified light rust, discoloration, flaking, and minor corrosion. There were no detrimental flaws or significant degradation noted during the examination. The areas did not affect the structural integrity or leak tightness of the SCV. The areas were cleaned and recoated in accordance with site procedures. A VT-3 preservice examination was performed following reapplication of coatings to satisfy the requirements of IWE-2200(g).

In 2007, a visual examination of the SQN, Unit 1 SCV exterior surface identified light corrosion deposits. There were no detrimental flaws or significant degradation noted during the examination. The areas did not affect the structural integrity or leak tightness of the SCV. The areas were cleaned and recoated in accordance with site procedures. A VT-3 preservice examination was performed following reapplication of coatings to satisfy the requirements of IWE-2200(g).

The conditions noted during the second interval IWE examinations consist of some flaking and discoloration of coatings along with rust, unpainted areas, arc strikes, scratches, small gouges, surface corrosion, scrapes, dings, and minor pitting which did not affect the leak tightness or structural integrity of the containment boundary. No detrimental flaws have been observed to date.

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NRC Information Notice 2004-09, "Corrosion of Steel Containment and Containment Liner," (Reference 17), discussed operating experience related to the potential for corrosive conditions at the junction of the metal liner and interior concrete floor. The SQN, Units 1 and 2, SCVs each have a moisture barrier that is examined in accordance with IWE-2500-1, Category E-A, Item E1.30. Examinations were conducted in 2000, 2003, 2006, 2011, and 2014 for SQN, Unit 2. and 2000, 2004, 2006, and 2012 for SQN, Unit 1.

In each of the exams, areas were noted where the moisture barrier was not completely adhered to the concrete interface. The moisture barrier was removed and reapplied in the subject areas. With the moisture barrier exposed, a VT-3 visual examination was performed to satisfy the requirements of IWE-2500(b). No detrimental flaws or significant degradation of the SCV liner was noted during these visual exams.

In 2011, four locations were identified in the SQN, Unit 2 SCV where the moisture barrier was not completely adhered to the concrete interface. Supplemental UT thickness measurements were taken to validate the remaining wall thickness in the excavated area. The thickness was above the nominal wall thickness and no further actions were required.

SQN, Units 1 and 2, have completed requirements of the first period, second 10-year interval Containment IWE ISI Program. The second 10-year interval IWE examination requirements use the 2001 Edition, 2003 Addenda, of ASME Section Xl as modified by the 10 CFR 50.55a(b) limitations for both units. The examinations are based on Category E-A and are visual (General, VT-3, and VT-1) examinations based on ASME Code or 10 CFR rule requirements. Currently, SQN, Units 1 and 2 have no areas requiring augmented examination per the requirements of Category E-C.

4.4.2 IWL Examinations Subsection IWL provides the examination and exemption requirements for Class CC components. The Class CC equivalent components at SQN, i.e., the base slab and metal liner, are exempt from examination based on the exemptions of IWL-1220(b) and IWE-1220(b). Class CC components exempt from examination are:

1. Tendon end anchorages that are inaccessible, subject to the requirements of IWL-2521.1

2. Portions of the concrete surface that are covered by the liner, foundation material, or back fill, or are otherwise obstructed by adjacent structures, components, parts, or appurtenances There are no tendons associated with SQN, Units 1 and 2, SCVs. The structural base slab and metal liner are covered with concrete, which forms the reactor building floor and results in these components being inaccessible for examination.

4.5 Deficiencies Identified Consistent with the guidance provided in NEI 94-01, Revision 3-A, Section 9.2.3.3, abnormal degradation of the primary containment structure identified during the conduct Page E1-13 of 27

of IWE/IWL program examinations or at any other times would be entered into the corrective action program for evaluation to determine the cause of the degradation and to initiate appropriate corrective actions.

4.6 Plant Specific Confirmatory Analysis 4.6.1 Methodology A plant-specific risk assessment was performed to assess the effect on plant risk of extending the SQN, Units 1 and 2, ILRT surveillance intervals from the current 10 years to 15 years. The assessment is included as Enclosure 2 to this letter. This assessment followed the guidance of NEI 94-01, Revision 3-A, the methodology described in EPRI Report 1018243, and the NRC regulatory guidance outlined in Regulatory Guide 1.174 on the use of PRA and risk insights in support of a license amendment request (LAR) for changes to a plants licensing basis. In addition, the methodology used for the Calvert Cliffs Nuclear Power Plant (CCNP) to estimate the likelihood and risk implication of undetected corrosion-induced leakage of steel containment liners for the additional window of vulnerability from extending the ILRT interval was used to estimate the conditional containment failure probability and its effect on the Large, Early Release Frequency (LERF) and the estimated population dose. The estimated values determined by using the CCNP methodology were used as part of an analysis to characterize the sensitivity of the risk impact results from extending the ILRT interval.

Prior to issuance, Revision 5 of the SQN PRA model was subjected to a full peer review against American Society of Engineers/American National Standard (ASME/ANS) RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 18). The peer review was conducted in accordance with NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance" (Reference 19),

including the makeup of the review team. All technical elements of the ASME/ANS Standard were assessed for applicability to SQN's Internal Events (including internal flooding) PRA. As such, the SQN PRA was evaluated against all applicable supporting level requirements (SLRs). All Facts & Observations (F&Os) identified during the peer review were either incorporated into the PRA prior to the Revision 5 model being issued, or resolved as described in Enclosure 3 to this letter.

The SQN PRA has a configuration control program consistent with the requirements of ASME/ANS RA-Sb-2013, Section 1-5.2. As such, the PRA model has been updated; however, no upgrades have been performed including the current model of record (MOR), Revision 8 (August 5, 2014). The Revision 8 PRA MOR for internal events (including internal flooding) was used in the ILRT interval extension analysis.

In the SE issued by NRC letter dated June 25, 2008, the NRC concluded that the methodology in EPRI Report No. 1018243, Revision 2-A of 1009325, is acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.2 of the SE. The following table addresses each of the four limitations and conditions for use of EPRI Report 1009325, Revision 2-A. These limitations and conditions were incorporated into EPRI Report 1018243 (Reference 13).

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From Safety Evaluation (SE) of EPRI - 1009325 SQN Response

1. The licensee submits documentation indicating SQN, Units 1 and 2, PRA technical adequacy is that the technical adequacy of their PRA is addressed in Section 4.6.2.

consistent with the requirements of RG 1.200 relevant to the ILRT extension.

2. The licensee submits documentation indicating EPRI Report No. 1018243, Revision 2-A of that the estimated risk increase associated with 1009325, incorporates these population dose and permanently extending the ILRT surveillance CCFP acceptance guidelines. These guidelines interval to 15 years is small and consistent with have been used for the SQN, Units 1 and 2, the specific risk clarification provided in assessment.

Section 3.2.4.5 of the SE. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1%

of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in a previous one-time ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

3. The methodology in EPRI Report No. 1009325, EPRI Report No. 1018243, Revision 2-A of Revision 2, is acceptable except for the 1009325, incorporated the use of 100 La as the calculation of the increase in expected average leak rate for the pre-existing containment population dose (per year of reactor operation). large leakage rate accident case (accident case 3b).

In order to make the methodology acceptable, This value has been used in the SQN, Units 1 and 2, the average leak rate for the pre-existing specific risk assessment.

containment large leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

4. A LAR is required in instances where SQN, Units 1 and 2, do not rely on containment containment over-pressure is relied upon for overpressure to assure adequate ECCS pump net emergency core cooling system (ECCS) positive suction head following design basis performance. accidents. No additional risk analysis was needed for this consideration.

4.6.2 PRA Technical Adequacy NRC RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (Reference 20) describes an acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors. To demonstrate that the technical adequacy of the PRA used in an application is of sufficient quality, RG 1.200 states that the staff expects the following information to be submitted to the NRC. RG 1.200 also states that previously submitted documentation may be referenced if it is adequate for the subject submittal:

1. [Assurance that] the PRA model represents the as-designed or as-built, as-operated plant.

The SQN Revision 5 PRA was subjected to peer review in 2011. The Peer Review Team used RG 1.200, Revision 2; however, it is important to note SQN does not have PRA models for external events. The model does include internal flooding initiating events.

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The ASME Standard endorsed by RG 1.200 (i.e., ASME/ANS RA-Sa-2009) includes High Level Requirements (HLRs) that address the PRA with respect to representing the as-built, as-operated plant, i.e., HLR-DA-B and HLR-MU-B. There are six SLRs for those sub-elements. The Peer Review Team characterization of these SLRs is that the SQN PRA meets the supporting requirements with a Capability Category (CC) I - III.

Furthermore, to ensure the PRA model is maintained to represent the as-built, as-operated plant, TVA Procedure NEDP-26, "Probabilistic Risk Assessment (PRA),"

Reference 21) states, "Various information sources shall be monitored by the Corporate/Site PRA Specialist on an ongoing basis to determine changes or new information that will affect the model, model assumptions, or quantification.

Information sources include Operating Experience, Technical Specification changes, plant modifications, Maintenance Rule changes, engineering calculation revisions, procedure changes, industry studies, NRC information and [Problem Evaluation Reports]."

2. Identification of permanent plant changes (such as deign or operational practices) that have an impact on those things modeled in the PRA but have not been incorporated in the baseline PRA model. If a plant change has not been incorporated, the licensee provides a justification of why the change does not impact the PRA results used to support the application. This justification should be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application decision were not adversely impacted (remained the same).

TVA procedure NPG-SPP-09.3, "Plant Modifications and Engineering Change Control," (Reference 22) governs the process for making changes across the TVA fleet. This procedure includes checklists that require the engineer to assess the potential effect on PRA criteria. If there is a "yes" response to any of the questions, an interface with the PRA group is required, and a review of the proposed design modification is performed in accordance with the PRA Program. As of the date of this submittal, there are no outstanding plant changes that necessitate a change to the SQN MOR dated August 5, 2014.

3. Documentation that the parts of the PRA required to produce the results used in the decision are performed consistently with the standard as endorsed in the appendices of this regulatory guide [RG 1.200, Revision 2]. If a requirement of the standard (as endorsed in the appendix to this guide [RG 1.200, Revision 2]) has not been met, the licensee is to provide a justification of why it is acceptable that the requirement has not been met. This justification should be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application were not impacted (remained the same).

The Peer Review Team determined that of the 325 supporting requirements (SRs),

313 were applicable to the SQN PRA. Of these, all but 19 were met at CC II or higher (i.e., CC-II, CC-III, CC-I/II, CC-II/III). The 19 SRs are further divided as 11 not met, and 8 met at CC-I.

The proposed resolutions for the F&Os associated with these 19 SRs were resolved and incorporated as recommended by the Peer Team with a few exceptions as described as follows for 1-15 (Finding), 4-1 (Suggestion) and 4-13 (Suggestion).

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One aspect of F&O 1-15 (SRs AS-B1, AS-A10, and SC-B3) regarded not having explicit treatment of Station Blackout (SBO) in the accident scenario analysis (e.g., event trees). This aspect was discussed in the TVA response to NRC Request for Additional Information (RAI) 7.a.viii.4 in TVA letter dated July 17, 2013 (Reference 23). As stated in the TVA response, inclusion of the SBO system failures post power recovery has a negligible effect on the Core Damage Frequency (CDF);

therefore, there would be insignificant effect on the calculations to risk-informed surveillance frequencies. The treatment of F&O 1-15 was specifically addressed in the TVA responses to NRC RAIs PRA-KNH-008 and PRA-KNH-009 associated with the review of the proposed amendment request converting the SQN, Units 1 and 2, TS to the improved Standard TS of NUREG-1431.

F&O 4-1 proposes revising tables in the flooding analysis. However, the flooding analysis already had the proposed information in a table and the recommendation would have added an unnecessary redundancy. The treatment of F&O 4-1 was specifically addressed in the TVA responses to NRC RAIs PRA-KNH-008 and PRA-KNH-009 associated with the review of the proposed amendment request converting the SQN, Units 1 and 2, TS to the improved Standard TS of NUREG-1431.

F&O 4-13 with respect to SR DA-C8 suggests using plant-specific operational records to determine the time that components are configured in their standby status. The change in risk associated with a change in the periodicity of surveillance frequencies is not expected to be affected by the use of split-fractions for systems with a normally operating train/pump, and another in standby. The treatment of F&O 4-13 was specifically addressed in the TVA responses to NRC RAIs PRA-KNH-008 and PRA-KNH-009 associated with the review of the proposed amendment request converting the SQN, Units 1 and 2, TS to the improved Standard TS of NUREG-1431.

4. A summary of the risk assessment methodology used to assess the risk of the application, including how the base PRA model was modified to appropriately model the risk impact of the application and results. (Note that this is the same as that required in the application-specific regulatory guides.)

The SQN PRA approach uses computer aided fault tree analysis (CAFTA) and provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired consequences (e.g., core damage and/or large early release) and their frequencies, and is comprised of specific technical elements (e.g., data, HRA, initiators) in performing the quantification. This evaluation is to document the PRA technical adequacy to support the proposed risk-informed application to permanently extend the containment Type A ILRT. The analysis to support the change in risk used the existing PRA model without having to make modifications to the model.

The methodology, which is described fully in the calculation provided in Enclosure 2 to this letter, employed to assess the risk of this application was performed in accordance with the industry guidance for implementation of the performance-based option of 10 CFR 50, Appendix J and the EPRI risk-impact assessment of extended integrated containment leak-rate testing intervals.

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5. Identification of the key assumptions and approximations relevant to the results used in the decision-making process. Also, include the peer reviewers assessment of those assumptions. These assessments provide information to the NRC staff in their determination of whether the use of these assumptions and approximations is appropriate for the application, or whether sensitivity studies performed to support the decision are appropriate.

The Peer Review Team characterized the internal events SRs associated with key assumptions and approximations and graded these at CC II or I-III. Technical elements included initiating events (IE), accident sequence analysis (AS), success criteria (SC), data analysis (DA), systems analysis (SY), human reliability (HRA),

quantification (QU) and large early release (LE). Additionally the rare event approximation received a CC I-III in the quantification (QU) technical element.

The proposed application uses the accident sequence results from the Level 2 analysis. All assumptions associated with the ILRT calculation are identified and discussed in the calculation and does not affect the PRA modeling, but rather the results from the PRA.

Assumptions associated with the Level 2 analysis discussed by the Peer Review Team included over-conservatisms such as use of steam generator Power Operated Relief Valves (PORVs). The peer review F&Os have all been resolved; therefore, there is no effect on the proposed application.

In addition to LERF, the acceptance criteria for the proposed application include CCFP, the change in the CCFP (CCFP), and the change in population dose. Only the metrics associated with LERF are considered by the PRA model.

6. A discussion of the resolution of the peer review (or self-assessment, for peer reviews performed using the criteria in NEI 00-02) findings and observations that are applicable to the parts of the PRA required for the applications. This decision should take the following forms:

a. a discussion of how the PRA model has been changed

b. a justification in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application decision were not adversely impacted (remained the same) by the particular issue.

The Peer Review Team reviewed an initial revision of the PRA model using CAFTA.

The CAFTA model replaced the RiskMan model previously used at SQN. Because the Team reviewed a draft model, their recommendations did not make a change to the model, but rather supplemented the completion of the model to a Revision 0 status.

Standard sensitivity analyses were performed on the following:

• All Human Error Probabilities set to their 5th percentile value

• All Human Error Probabilities set to their 95th percentile value

• All CCF probabilities set to their 5th percentile value

• All CCF probabilities set to their 95th percentile value

• All maintenance terms set to zero Page E1-18 of 27

Model specific sensitivity analyses were performed on the following:

• Grid reliability and loss of offsite power (LOOP) modeling

• LOOP frequency, recovery, consequential LOOP, timing

• Post-LOOP equipment recovery

• Credit for non-safety related equipment

• Battery depletion

7. The standards or peer review process documents may recognize different capability categories or grades that are related to level of detail, degree of plant specificity, and degree of realism. The licensees documentation is to identify the use of the parts of the PRA that conform to capability categories or grades lower than deemed required for the given application (Section 1-3 of ASME/ANS RA-Sa-2009).

The proposed application requires a quality level of CC II. The Peer Review Team stated the SQN PRA meets the ASME/ANS Standard, and concluded the PRA uses processes and tools that are at the state of the technology and generally consistent with CC II.

4.6.3 Conclusion of Plant-Specific Risk Assessment Results The SQN, Units 1 and 2, risk assessment associated with extending the ILRT surveillance interval from three Type-A ILRT tests in 10 years to one Type A ILRT test in 15 years is small, consistent with other generalized and specific industry analysis on this same subject. Details of the SQN, Units 1 and 2, risk assessment are contained in Enclosure 2 to this letter. The SQN-specific results for extending the ILRT surveillance interval from the current 10 years to 15 years are summarized below. In the discussions that follow the maximum results are provided for LERF, CCFP, and Dose. The SQN, Unit 2, results bound the SQN, Unit 1, results for all three parameters, therefore only the SQN, Unit 2, values are presented in the discussions. However, Table 4.6.3-1, below, provides the data for both units.

1. Regulatory Guide 1.174 (Reference 12) 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 CDF; therefore, there is no change to the plant CDF as a result of implementing this proposed change to the licensing basis. The guidance provided in Regulatory Guide 1.174 describes a small change in risk for LERF as less than 1.0E-06/reactor-year (rx-yr), if it can be reasonably shown that the total LERF is less than 1.0E-05/rx-yr. For SQN, the analysis included the estimated contribution from external events in addition to the internal events analysis. The maximum LERF for SQN was estimated to be 5.90E-07/yr, and the total LERF was estimated to be 6.84E-06/yr. Both results are within the acceptable bands for a small change in risk as defined by Regulatory Guide 1.174. Table 4.6.3-1, below, provides the results for SQN, Units 1 and 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 three tests in 10 years as compared to the current licensing basis (CLB) of one test in 10 years, and the proposed licensing basis (PLB) of one test in 15 years.

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2. The maximum calculated increase in the CCFP increase for SQN, Units 1 and 2, from the OLB to the PLB is 0.842%. EPRI Report 1018243 characterizes an increase in the CCFP of 1.5% as small. Therefore, this increase is determined to be small. Table 4.3.6-1, below, provides the detailed results for SQN, Units 1 and 2, for the %CCFP change for the CLB and PLB compared to the OLB.

3. The maximum calculated population dose for SQN, Units 1 and 2, from the OLB to the PLB is 6.74E-02 person-rem/yr, corresponding to increase of 0.57%. This value is based on internal events only. EPRI Report 1018243 (Reference 13) states that a very 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 ILRT interval extension to 15 years. Therefore, this increase is determined to be very small. Table 4.3.6-1, below, provides the detailed results for SQN, Units 1 and 2, for the total change in dose and the percentage change for the CLB and PLB compared to the OLB.

4. As noted in the table in Section 4.6.1, SQN, Units 1 and 2, do not credit containment overpressure to satisfy net positive suction head requirements for containment spray and low-head safety injection in recirculation mode during LOCAs.

Table 4.6.3-1 PRA Results Table Unit 1 Acceptable for Metric Value Acceptance Criteria Application?

LERFIE-Total 2.98E-06/yr

<1.0E-05/rx-yr Yes LERFTotal(IE & EE) 6.68E-06/yr LERFTotal(1-in-3 1-in-10) 3.39E-07/yr

<1.0E-06/rx-yr Yes LERFTotal(1-in-3 1-in-15) 5.81E-07/yr CCFP(1-in-3 1-in-10) 0.490%

1.5% Yes CCFP(1-in-3 1-in-15) 0.841%

DOSE(1-in-3 1-in-10) 3.75E-02 per-rem/yr DOSE(1-in-3 1-in-15) 6.43E-02 per-rem/yr <1.0 person-rem/yr or <1% of total Yes

%DOSE(1-in-3 1-in-10) 0.33% dose, whichever is less restrictive.

%DOSE(1-in-3 1-in-15) 0.57%

Unit 2 LERFIE-Total 3.05E-06/yr

<1.0E-05/rx-yr Yes LERFTotal(IE & EE) 6.84E-06/yr LERFTotal(1-in-3 1-in-10) 3.44E-07/yr

<1.0E-06/rx-yr Yes LERFTotal(1-in-3 1-in-15) 5.90E-07/yr CCFP(1-in-3 1-in-10) 0.490%

1.5% Yes CCFP(1-in-3 1-in-15) 0.842%

DOSE(1-in-3 1-in-10) 3.93E-02 per-rem/yr DOSE(1-in-3 1-in-15) 6.74E-02 per-rem/yr <1.0 person-rem/yr or <1% of total Yes

%DOSE(1-in-3 1-in-10) 0.33% dose, whichever is less restrictive.

%DOSE(1-in-3 1-in-15) 0.57%

Page E1-20 of 27

5.0 REGULATORY EVALUATION

5.1 Applicable Regulatory Requirements and Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing, and reporting requirements for each type of test. RG 1.163 was developed to endorse NEI 94-01, Revision 0 with certain modifications and additions.

The adoption of the Option B performance-based containment leakage rate testing for Type A testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviews "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The previous change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type A test frequency is not expected to result in an increase in containment leakage.

NEI 94-01, Revision 3-A, describes an approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. The document incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A and Type C intervals to 15 years and 75 months, respectively. NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate test frequencies. In the SE issued by NRC letter dated June 8, 2012, the NRC concluded that NEI 94-01, Revision 3-A, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their TS with regard to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.0 of the SE.

EPRI Report 1009325, Revision 2, provides a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94-01, Revision 3-A, states that a plant-specific risk impact assessment should be performed using the approach and methodology described in EPRI Report 1009325, Revision 2, for a proposed extension of the ILRT interval to 15 years. In the SE issued by NRC letter dated June 25, 2008, the NRC concluded that the methodology in EPRI Report 1009325, Revision 2, is acceptable for reference by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of that SE.

Based on the considerations 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 continue to be conducted in accordance with the site licensing Page E1-21 of 27

basis, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

In conclusion, TVA has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements/criteria.

5.2 Precedent TVA has determined that this request is similar to the following license amendments, which have been approved by the NRC:

1. "Nine Mile Point Nuclear Station, Unit 2 - Issuance of Amendment Re: Extension of Primary Containment Integrated Leakage Rate Testing Interval (TAC No. ME1650),"

approved March 30, 2010 (ADAMS Accession Number ML100730032), Type A test interval only.

2. "Arkansas Nuclear One, Unit No.2 - Issuance of Amendment Re: Technical Specification Change to Extend the Type A Test Frequency to 15 Years (TAC No. ME4090)," approved April 7, 2011 (ADAMS Accession Number ML110800034), Type A test interval only.

3. "Palisades Nuclear Plant - Issuance of Amendment to Extend the Containment Type A Leak Rate Test Frequency to 15 Years (TAC No. ME5997)," approved April 23, 2012 (ADAMS Accession Number ML120740081), Type A test interval only.

4. "Virgil C. Summer Nuclear Station, Unit 1 - Issuance of Amendment Extending Integrated Leak Rate Test Interval (TAC No. MF1385)," approved February 5, 2014 (ADAMS Accession Number ML13326A204) Type A test interval only.

5. "Surry Power Station, Units 1 and 2 - Issuance of Amendment Regarding the Containment Type A and Type C Leak Tests (TAC Nos. MF2612 and MF2613),"

approved July 3, 2014 (ADAMS Accession Number ML14148A235), Type A and Type C test intervals.

5.3 No Significant Hazards Consideration The Tennessee Valley Authority (TVA) proposes to revise the current licensing basis of Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, by revising the SQN, Units 1 and 2, Technical Specifications (TS) 6.8.4.h, "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," 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 (ILRT) interval from 10 years to 15 years and the Type C local leakage rate testing (LLRT) intervals from 60 months to 75 months.

TVA has concluded that the changes to SQN, Units 1 and 2, TS 6.8.4.h do not involve a significant hazards consideration. TVAs conclusion is based on its evaluation in Page E1-22 of 27

accordance with 10 CFR 50.91(a)(1) of 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 consequence of an accident previously evaluated?

Response: No.

The proposed revision to TS 6.8.4.h changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR Part 50, Appendix J, Option B, LLRT). The current type A test interval of 10 years would be extended to 15 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident because research documented in NUREG-1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. NUREG-1493 concluded that reducing the Type A testing frequency to one per twenty years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The last Type A test (performed October 27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2) shows leakage to be below acceptance criteria, indicating a very leak tight containment. Inspections required by the ASME Code Section Xl (Subsections IWE and IWL) and Maintenance Rule monitoring (10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants") are performed in order to identify indications of containment degradation that could affect that leak tightness. Types B and C testing required by TSs will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test interval extension will not represent a significant increase in the consequences of an accident.

The proposed amendment involves changes to the SQN, Units 1 and 2, 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the units are operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the SQN, Units 1 and 2, performance-based leakage testing program. 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 potential consequences of extending Page E1-23 of 27

the ILRT interval from 10 years to 15 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year resulting from design basis accidents was estimated to be very small, and the increase in the LERF resulting from the proposed change was determined to be within the guidelines published in NRC RG 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. TVA has determined that the increase in CCFP due to the proposed change would be very small.

Based on the above discussions, the proposed changes do not involve an increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed revision to TS 6.8.4.h changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR Part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (performed October 27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The proposed extension to Type A and Type C test intervals does not create the possibility of a new or different type of accident because there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed revision to TS 6.8.4.h changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR Part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test

((performed October 27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The proposed extension to Type A testing will not significantly reduce the margin of safety. NUREG-1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, generic study of the effects of extending containment leakage testing, found that a 20 year extension to Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1% to the individual risk and that the decrease in Type A Page E1-24 of 27

testing frequency would have a minimal effect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the ASME Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) will further reduce the risk of a containment leakage path going undetected.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the SQN, Units 1 and 2, performance-based leakage testing program, and establishes a 15-year interval for the performance of the primary containment ILRT and a 75-month interval for Type C testing. The amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the 10 CFR Part 50, Appendix J Testing Program Plan, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant safety analyses is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, B, and C containment leakage tests will continue to be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A.

Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by an ILRT. This ensures that evidence of containment structural degradation is identified in a timely manner.

Furthermore, a risk assessment using the current SQN, Units 1 and 2, PRA model concluded that extending the ILRT test interval from 10 years to 15 years results in a very small change to the SQN, Units 1 and 2, risk profile.

Accordingly, the proposed changes do not involve a significant reduction in a margin of safety.

5.4 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 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. 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 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 Page E1-25 of 27

statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 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 (ADAMS Accession No. ML12221A202).

2. TVA Letter to NRC, "Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10)," dated November 22, 2013 (ADAMS Accession No. ML13329A717).

3. 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."

4. NRC Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 (ADAMS Accession No. ML003740058).

5. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, dated July 26, 1995 (ADAMS Legacy Library Accession No. 9510200180).

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

7. Electrical Power Research Institute (EPRI) Report 104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994.

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

9. Letter from M. J. Maxin, NRC, to J. C. Butler, 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 (ADAMS Accession No. ML081140105.

10. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2-A, dated October 2008 (ADAMS Accession No. ML100620847).

11. EPRI Report 1009325, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," Revision 2, dated August 2007 (ADAMS Accession No. ML072970208).

Page E1-26 of 27

12. 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 2, dated May 2011 (ADAMS Accession No. ML100910006).

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

14. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3, dated June 2011 (ADAMS Accession No. ML112920567).

15. Letter from Sher Bahadur, NRC, to Mr. Biff Bradley, 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 (ADAMS Accession No. ML121030286).

16. NRC Regulatory Issue Summary (RIS) 2008-27, "Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50," dated December 8, 2008 (ADAMS Accession No. ML080020394).

17. NRC Information Notice 2004-09, "Corrosion of Steel Containment and Containment Liner," dated April 27, 2004.

18. American Society of Mechanical Engineers/American National Standard (ASME/ANS) RA-S-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated February 2009.

19. NEI 00-02, " Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"

Revision A3, dated March 20, 2000 (ADAMS Accession No. ML003728023).

20. 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 (ADAMS Accession No. ML090410014).

21. TVA Procedure NEDP-26, "Probabilistic Risk Assessment (PRA), Revision 9, dated December 13, 2003.

22. TVA Procedure NPG-SPP-09.3, "Plant Modifications and Engineering Change Control," Revision 17, dated June 2, 2014.

23. TVA Letter, "Response to NRC Request for Additional Information Regarding the Environmental Review of the Sequoyah Nuclear Plant, Units I and 2, License Renewal Application (TAC Nos. MF0057 and MF0058)," dated July 17, 2013 (ADAMS Accession No. ML13227A003).

Page E1-27 of 27

ATTACHMENT 1 Proposed TS Changes (Mark-Ups) for SQN, Units 1 and 2

Proposed TS Changes (Mark-Ups) for SQN, Unit 1

h. 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. Visual examination and testing, including test intervals and extensions, shall be in accordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing NEI 94-01, Industry instructions and the following:

Guideline for Performance-Based Option of 10 CFR 50, BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves Appendix J," Revision 3-A, July 2012, and Section 4.1, that are sealed with fluid from a seal system may be excluded, subject to the provisions of "Limitations and Conditions Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal for NEI TR 94-01, Revision 2," of the NRC system and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacity Safety Evaluation Report in is adequate to maintain system pressure (or fluid head for the containment spray system and NEI 94-01, Revision 2-A, dated October 2008, RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are: 0.60 La for the combined Type B and Type C tests, and: 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is : 0.05 La when tested at Pa.

2. For each door, leakage rate is: 0.01 La when pressurized to 6 psig for at least two minutes.

c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type Band C tests.

2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test.

The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

i. Configuration Risk Management Program (DELETED)

April 13, 2009 SEQUOYAH - UNIT 1 6-10a Amendment No. 217, 241, 281, 287, 323

ADMINISTRATIVE CONTROLS

i. Configuration Risk Management Program (DELETED)

j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of TSs.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

k. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full February 23, 2006 SEQUOYAH - UNIT 1 6-11 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306

Proposed TS Changes (Mark-Ups) for SQN, Unit 2 ADMINISTRATIVE CONTROLS 6.8.4 f. Radioactive Effluent Controls Program (Cont.)

of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:

1. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

2. For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, lodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and

10) Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency.

g. Radiological Environmental Monitoring Program (DELETED)

h. 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. Visual examination and testing, including test intervals and extensions, shall be in accordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing instructions and the following:

NEI 94-01, Industry Guideline for Performance-Based Option of 10 CFR 50, BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves Appendix J," Revision 3-A, that are sealed with fluid from a seal system may be excluded, subject to the provisions of July 2012, and Section 4.1, Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal "Limitations and Conditions for NEI TR 94-01, system and valves are pressurized to at least 1.10 Pa(13.2 psig) and the seal system Revision 2," of the NRC capacity is adequate to maintain system pressure (or fluid head for the containment spray Safety Evaluation Report in system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least NEI 94-01, Revision 2-A, dated October 2008, 30 days.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

April 13, 2009 SEQUOYAH - UNIT 2 6-9 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 265, 272, 276, 315

ADMINISTRATIVE CONTROLS The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is 0.05 La when tested at Pa.

2) For each door, leakage rate is 0.01 La when pressurized to 6 psig for at least two minutes.

c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type B and C tests.

2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test.

The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

i. Configuration Risk Management Program (DELETED)

j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of these TSs.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

April 13, 2009 SEQUOYAH - UNIT 2 6-10 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 315

ATTACHMENT 2 Proposed TS Changes (Final Typed) for SQN, Units 1 and 2

Proposed TS Changes (Final Typed) for SQN, Unit 1

h. 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. Visual examination and testing, including test intervals and extensions, shall be in accordance with 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, with exceptions provided in the site implementing instructions and the following:

BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are: 0.60 La for the combined Type B and Type C tests, and: 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is : 0.05 La when tested at Pa.

2. For each door, leakage rate is: 0.01 La when pressurized to 6 psig for at least two minutes.

c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type Band C tests.

2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test.

The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

April 13, 2009 SEQUOYAH - UNIT 1 6-10a Amendment No. 217, 241, 281, 287, 323

ADMINISTRATIVE CONTROLS

i. Configuration Risk Management Program (DELETED)

j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of TSs.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

k. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full February 23, 2006 SEQUOYAH - UNIT 1 6-11 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306

Proposed TS Changes (Final Typed) for SQN, Unit 2 ADMINISTRATIVE CONTROLS 6.8.4 f. Radioactive Effluent Controls Program (Cont.)

of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:

1. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

2. For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, lodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and

10) Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency.

g. Radiological Environmental Monitoring Program (DELETED)

h. 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. Visual examination and testing, including test intervals and extensions, shall be in accordance with 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, with exceptions provided in the site implementing instructions and the following:

BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa(13.2 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

April 13, 2009 SEQUOYAH - UNIT 2 6-9 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 265, 272, 276, 315

ADMINISTRATIVE CONTROLS The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is 0.05 La when tested at Pa.

2) For each door, leakage rate is 0.01 La when pressurized to 6 psig for at least two minutes.

c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type B and C tests.

2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test.

The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

i. Configuration Risk Management Program (DELETED)

j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of these TSs.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

April 13, 2009 SEQUOYAH - UNIT 2 6-10 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 315

ATTACHMENT 3 Proposed improved TS Page Mark-Ups for SQN, Units 1 and 2

Proposed improved TS Page Mark-Up for SQN, Unit 1 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leakage Rate Testing Program

a. A program shall establish 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 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as NEI 94-01, Industry modified by the following exceptions:

Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, 1. Bypass leakage paths to the auxiliary building leakage from isolation July 2012, and Section 4.1, valves that are sealed with fluid from a seal system may be excluded, "Limitations and Conditions for NEI TR 94-01, subject to the provisions of Appendix J,Section III.C.3, when Revision 2," of the NRC Safety Evaluation Report in determining the combined leakage rate provided the seal system and NEI 94-01, Revision 2-A, valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal dated October 2008, system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

SEQUOYAH - UNIT 1 5.5-3 Amendment XXX

Proposed improved TS Page Mark-Up for SQN, Unit 2 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leakage Rate Testing Program

b. A program shall establish 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 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as NEI 94-01, Industry modified by the following exceptions:

Guideline for Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A,

2. Bypass leakage paths to the auxiliary building leakage from isolation July 2012, and Section 4.1, valves that are sealed with fluid from a seal system may be excluded, "Limitations and Conditions for NEI TR 94-01, subject to the provisions of Appendix J,Section III.C.3, when Revision 2," of the NRC determining the combined leakage rate provided the seal system and Safety Evaluation Report in NEI 94-01, Revision 2-A, valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal dated October 2008, system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

SEQUOYAH - UNIT 2 5.5-5 Amendment XXX

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 RISK IMPACT ASSESSMENT

Subject:

Application to Revise Technical Specification 6.8.4.h, "Containment Leakage Rate Testing Program," (SQN-TS-14-03)

Page E2-1 of 77

Page E2-2 of 77 NPG CALCULATION COVERSHEET / CTS UPDATE CALC ID ORG PLANT BRANCH NUMBER REV NUC Sequoyah MEB MDN 000 999 2014 000162 000 BUILDING ROOM ELEVATION COORD/AZIM FIRM N/A N/A N/A N/A N/A CATEGORIES N/A KEYWORDS (A-add, D-delete)

ACTION KEYWORD A/D KEYWORD (A/D)

A PROBABILISTIC A PSA A RISK A PRA A ILRT A CCFP A LERF CROSS-REFERENCES (A-add, D-delete)

ACTION XREF XREF XREF XREF XREF (A/D) CODE Plant TYPE NUMBER REV CTS ONLY UPDATES:

Following are required only when making keyword/cross reference CTS updates and page 1 of form NEDP-2-1 is not included:

PREPARER (PRINT NAME AND SIGN) DATE CHECKER (PRINT NAME AND SIGN) DATE PREPARER PHONE NO. EDMS ACCESSION NO.

TVA 40532 Page 2 of 2 NEDP-2-1 [8-4-2014]

Page E2-3 of 77

Page E2-4 of 77 Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 4

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE NPG Calculation Coversheet / CTS Update .............................................................................. 1 NPG Calculation Record of Revision ......................................................................................... 3 NPG CALCULATION TABLE OF CONTENTS .................................................................... 4 NPG Computer Input File Storage Information Sheet .......................................................... 10 1.0 Purpose/Background .................................................................................................. 11 2.0 References and Acronyms.......................................................................................... 12 2.1 References.................................................................................................................... 12 2.2 Acronyms ..................................................................................................................... 13 3.0 Assumptions ................................................................................................................ 14 4.0 Ground Rules .............................................................................................................. 15 5.0 Methodology ................................................................................................................ 16 5.1 Step 1 - Baseline Risk Determination........................................................................ 20 5.2 Step 2 - Develop the Baseline Population Dose Per Year ........................................ 22 5.3 Step 3 - Evaluate the Risk Impact (Bin Frequency and Population Dose) ............ 23 5.4 Step 4 - Evaluate the Change in LERF and CCFP .................................................. 23 5.4.1 Evaluate Risk Impact - Change in LERF ................................................................ 23 5.4.2 Evaluate Risk Impact - Change in CCFP ................................................................ 24 5.5 Step 5 - Evaluate the Sensitivity of the Results ........................................................ 24 5.5.1 Containment Overpressure........................................................................................ 24 5.5.2 External Events ........................................................................................................... 24 6.0 Inputs ........................................................................................................................... 25 6.1 Decomposition of LERF Frequency and EPRI Classification................................ 26 Page E2-5 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 5

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE 7.0 Calculation................................................................................................................... 37 7.1 Step 1 - Baseline Risk Determination ............................................................................................ 37 7.1.1 Class 2 - Large Containment Isolation Failures ........................................................................... 38 7.1.2 Class 7 - Severe Accident Phenomena ........................................................................................... 38 7.1.3 Class 8 - Containment Bypass (ISLOCA, SGTR) ........................................................................ 38 7.1.4 Calculation of the 3a Probability and Frequency......................................................................... 39 7.1.5 Calculation of the 3b Probability and Frequency ........................................................................ 39 7.1.6 Class 1 - Intact Containment.......................................................................................................... 40 7.2 Step 2 - Develop the Baseline Population Dose ............................................................................. 41 7.2.1 50-Mile Radius Population Density ............................................................................................... 41 7.2.2 NUREG/CR-4551 Off-Site Consequence (Person-Rem Estimates) ............................................42 7.2.3 Sequoyah Specific Off-Site Consequence (Person-Rem Estimates) ............................................44 7.3 Step 3 - Risk Impact Evaluation.................................................................................................... 47 7.3.1 Risk Impact - Once-in-10 Years Test Interval ......................................................................... 48 7.3.2 Risk Impact - Once-in-15 Years Test Interval ......................................................................... 49 7.3.3 Dose-Rate Increase and Percentile Increase ............................................................................. 50 7.3.3.1. Population Dose-Rate Calculations.......................................................................................... 51 7.4 Step 4 - LERF and CCFP Changes ............................................................................................... 54 7.4.1 LERF Determination ............................................................................................................. 54 7.4.2 Conditional Containment Failure Probability .......................................................................... 55 7.4.3 Summary LERF - CCFP ................................................................................................... 56 8.0 Sensitivity Analyses .................................................................................................... 57 8.1 Liner Corrosion............................................................................................................................... 57 8.1.1 Assumptions Used In the Corrosion Sensitivity Analysis ........................................................ 58 8.1.2 Differences in the Sequoyah Design from Calvert Cliffs ......................................................... 59 8.1.3 Base Case Risk Assessment ..................................................................................................... 60 8.1.4 Likelihood of Non-Detected Containment Leakage and LERF Impact ...................................62 8.1.5 Corrosion Impact on CCFP ...................................................................................................... 63 8.1.6 Summary of Base Case and Corrosion Sensitivity Cases ........................................................ 64 8.1.7 Liner Corrosion Sensitivity Conclusion ................................................................................... 68 9.0 Evaluation of External Events ................................................................................... 68 Page E2-6 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE 9.1 Seismic Analysis .......................................................................................................... 68 9.2 Internal Fires Analysis ............................................................................................... 69 9.3 External Events Contribution ................................................................................... 70 9.3.1 External Events Contribution to CDF ...................................................................... 70 9.3.2 External Events Contribution to LERF ................................................................... 70 9.3.3 External Events Contribution to LERF for ILRT Interval Extension .................. 71 10.0 Results/Conclusion...................................................................................................... 74 10.1 Results Discussion - LERF ........................................................................................ 74 10.2 Results Discussion - CCFP ........................................................................................ 75 10.3 Results Discussion - Population Dose ....................................................................... 75 List of Tables Table 1 Detailed Description of EPRI Accident Classes ............................................... 18 Table 2 EPRI Release Classes (Containment Failure Classifications) ......................... 25 Table 3 Sequoyah Release Categories and EPRI Mapping .......................................... 26 Table 4 Decomposition of Sequoyah LERF Frequency and EPRI Classification ...... 26 Table 5 EPRI Accident Class Frequencies ..................................................................... 37 Table 6 50-Mile Radius Population Density ................................................................... 41 Table 7 Reported Person-Rem Estimates for Sequoyah Source Term Groups .......... 42 Table 8 Calculation of the Sequoyah Population Dose Risk at 50-Miles ..................... 43 Table 9 U1 - Baseline Dose Calculation (Without 3a & 3b) ......................................... 44 Table 10 U2 - Baseline Dose Calculation (Without 3a & 3b) ......................................... 45 Table 11 U1 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution) ... 46 Page E2-7 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 7

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE Table 12 U2 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution) ... 46 Table 13 U1 - Testing Once-in-10 Years Risk Profile ...................................................... 48 Table 14 U2 - Testing Once-in-10 Years Risk Profile ...................................................... 49 Table 15 U1 - Testing Once-in-15 Years Risk Profile ...................................................... 49 Table 16 U2 - Testing Once-in-15 Years Risk Profile ...................................................... 50 Table 17 U1 - Class 1 PDR Increase Due to Extended Type A ILRT Intervals............ 51 Table 18 U1 - Class 3a PDR Increase Due to Extended Type A ILRT Intervals.......... 51 Table 19 U1 - Class 3b PDR Increase Due to Extended Type A ILRT Intervals ......... 51 Table 20 U1 -Total PDR Increase Due to Extended Type A ILRT Intervals ............... 52 Table 21 U2 - Class 1 PDR Increase Due to Extended Type A ILRT Intervals ........... 53 Table 22 U2 - Class 3a PDR Increase Due to Extended Type A ILRT Intervals ......... 53 Table 23 U2 - Class 3b PDR Increase Due to Extended Type A ILRT Intervals ......... 53 Table 24 U2 Total PDR Increase Due to Extended Type A ILRT Intervals ................. 53 Table 25 Unit-1 Summary LERF - CCFP ................................................................... 56 Table 26 Unit-2 Summary LERF - CCFP ................................................................... 57 Table 27 SQN Liner Corrosion Base-Case Risk Assessment .......................................... 60 Table 28 Unit-1 Increase in LERF/yr................................................................................ 62 Table 29 Unit-2 Increase in LERF/yr................................................................................ 63 Table 30 Unit-1 Summary of Base Case and Corrosion Sensitivity Cases .................... 66 Table 31 Unit-2 Summary of Base Case and Corrosion Sensitivity Cases .................... 67 Table 32 External Events Contribution to Risk for ILRT Interval Extension ............. 72 Table 33 U1 - Upper Bound on All LERF ........................................................................ 73 Page E2-8 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE Table 34 U2 - Upper Bound on All LERF Contributors ................................................. 73 Table 35 Acceptance Criteria ............................................................................................ 74 Table 36 Results Table and Applicability Determination ............................................... 75 List of Equations Equation 1 Calculation of the Class 2 Frequency ................................................................ 38 Equation 2 Calculation of the Class 7 Frequency ................................................................ 38 Equation 3 Calculation of the Class 8 Frequency ................................................................ 38 Equation 4 Calculation of the Class 3a Probability ............................................................. 39 Equation 5 Calculation of the Class 3a Failure Frequency ................................................ 39 Equation 6 Calculation of the Class 3b Failure Probability ............................................... 40 Equation 7 Calculation of the Class 3b Failure Frequency ................................................ 40 Equation 8 Calculation of the Class 1 Frequency ................................................................ 40 Equation 9 Calculation of the Adjusted Class 1 Frequency ............................................... 40 Equation 10 Percent Increase in Total Population Dose-Rate (PDR) .............................. 52 Equation 11 LERF Determination for Class 3b .............................................................. 54 Equation 12 Change in CCFP .............................................................................................. 55 Equation 13 %Change CCFP .............................................................................................. 56 Equation 14 Total Likelihood of Non-Detected Containment Leakage ........................... 62 Equation 15 Liner Corrosion Non-LERF Frequency........................................................ 62 Equation 16 Liner Corrosion - Increase in LERF ............................................................ 62 Equation 17 Increase in CCFP Due to Increase in Flaw Likelihood ............................... 63 Equation 18 CCFP Increase Due to Corrosion................................................................ 64 Page E2-9 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN 000 999 2014 000162 Revision: 0 TABLE OF CONTENTS SECTION TITLE PAGE Equation 19 External Events Contribution to CDF .......................................................... 70 Equation 20 External Events Impact on the Class 3b Frequency .................................... 70 Equation 21 EE Impact on the Class 3b Frequencies for Extended ILRT Intervals ..... 71 Equation 22 External Events Contribution to a Change in LERF .................................. 72 Equation 23 Total External Events Impact on LERF from Extending ILRT Interval.. 72 TVA 40710 [10-2008] Page 2 of 2 NEDP-2-3 [10-20-2008]

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document MDN 000 999 2014 000162 Rev. 000 Plant: SQN

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Electronic storage of the input files for this calculation is not required. Comments:

Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file.

(Any retrieved file requires re-verification of its contents before use.)

All electronic files used in the creation of this calculation and the electronic version of this calculation are stored under FileKeeper number 324720.

Microfiche/eFiche TVA 40535 [10-2008] Page 1 of 1 NEDP-2-6 [10-20-2008]

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Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 11

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT 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 leakage testing which allows licensees with acceptable test performance history to extend surveillance intervals. At that time, provisions were made for extending ILRT frequency from the original licensing basis (OLB) of three-in-ten years to once-in-10 years, supported by the NRCs assessment (NUREG-1493)[7] that stated there is an imperceptible increase in risk associated with ILRT intervals up to twenty years.[5] During the early 2000s most licenses applied for one-time extensions to once-in-15 years, including Sequoyah.

Integrated leak-rate testing (ILRT) is the only method capable of detecting all existing leaks in the reactor containment system, which 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.[7]

The purpose of this analysis is to provide a risk assessment of permanently extending the Current Licensing Basis (CLB) allowed Containment Type A Integrated Leak Rate Test (ILRT) interval from once-in-10 years to the Proposed Licensing Basis (PLB) of once-in-15 years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the Sequoyah plant, which over the remaining life of the plant would reduce the total required number of Type A ILRTs.

Earlier assessments followed the guidance of EPRI TR-104285 that considered changes in local leak-rate testing and ILRT testing intervals.[3] 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).[2] This risk assessment follows the guidance from NEI 94-01[5], the methodology used in EPRI TR-1018243,[2] NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in RG 1.200,[13] and RG 1.174,[12] for risk insights in support of a request for a change to a plants licensing basis. Additionally, the Calvert Cliffs methodology to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval was used for the Sequoyah analysis.[14]

An earlier revision to 10CFR50, Appendix J (Option B) allowed individual plants to extend the Containment Type A ILRT surveillance testing requirement from three-in-ten years to once-in-10 years which is the current licensing basis for Sequoyah. The revised Type A frequency is based on an acceptable performance history.

This calculation evaluates the risk associated with the following ILRT intervals:

• 3 Years - Original Licensing Basis (OLB) test interval of three per 10 years

• 10 Years - Current Licensing Basis (CLB) test interval is once per 10 years

• 15 Years - Proposed Licensing Basis (PLB) test interval to once per 15 years Page E2-12 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 12

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval The risk analysis uses the Sequoyah PRA Level 2 m odel.[17] The release category and dose (person-rem) information is based on the approach suggest by the EPRI guidance.[1]

The NRC report on performance-based leak testing, NUREG-1493[7] 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 a determined that for a representative PWR (i.e., Surry) that containment isolation failures contribute less than 0.1 percent to the latent risk 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.

[7,§5.2.2.1]

Consequently, it is necessary to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for the Sequoyah Plant.

2.0 References and Acronyms 2.1 References

1. EPRI Report 1009325, December 2003, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals

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

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

4. NEI 94-01 Rev. 2, Industry Guideline for Implementation Performance-Based Option of 10CFR Part 50, Appendix J

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

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

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

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

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

Sequoyah, Unit 1

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

Sequoyah, Unit 1

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

12. Reg. Guide 1.174 Rev. 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis

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

14. 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.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No.

50-317, March 27, 2002. ML020920100

15. MDN-000-000-2010-0200 Rev. 3, Sequoyah Probabilistic Risk Assessment - Summary Notebook

16. MDN-000-000-2010-0206 Rev 3, Sequoyah Probabilistic Risk Assessment - Level 2 Analysis

17. Sequoyah-0-14-072, Sequoyah Level 2 Results in Support of ILRT Extension Request B45 140820 001

18. TVASEQ008-CALC-002 Rev. 0 Sequoyah WinMAACS Assessment of Severe Accident Consequences

19. Westinghouse, LTR-RAM-II-11-010, RG 1.200 PRA Peer Review Against the ASME/ANS Standard Requirements for the Sequoyah Nuclear Plant Probabilistic Risk Assessment

20. Sequoyah Nuclear Plant Individual Plant Evaluation of External Events, June 29, 1995

21. Surveillance Instruction, 1-SI-SLT-088-156.0, Containment Integrated Leak Rate Test

22. Surveillance Instruction, 1-SI-DXI-000-254.1, Containment Vessel Integrity Verification

23. SQS20211 Rev. 1, Evaluation of the Risk Significance of Decreased Containment Integrated Leak Test Frequency

24. Sequoyah Nuclear Power Plant, Updated Final Safety Analysis Report, Amendment 24

25. ADAMS ML13024A010, Sequoyah Nuclear Plant, Applicants Environmental Report Operating License Renewal Stage, Attachment E, Evaluation of SQN PRA Model, Section E.1.3 IPEEE Analysis

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

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

28. ML13329A718, SQN ITS Project Submittal/Supporting PRA Documentation

29. ML100270664, GI-199, Appendix A Seismic Core-Damage Frequency Estimates

30. ML100270691, GI-199, Appendix B Seismic Hazards Estimate

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

32. SQN IPEEE Response to Supplement 4 of GL-88-20, July 11, 1999

33. RIMS B87010829001, Sequoyah Evaluation of the Risk Significance of Decreased Containment Integrated Leak Rate Test Frequency, April 15, 2002 2.2 Acronyms 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 Page E2-14 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval CLB - Current Licensing Basis EPRI - Electric Power Research Institute F&O - Facts & Observations HCLPF - High-Confidence of Low-Probability of Failure ILRT - 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 SQN - Sequoyah Nuclear Plant SGTR - Steam Generator Tube Rupture STG - Source Term Group 3.0 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.[2]

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

3. The assumed maximum containment leakage (large) for Class 3b sequences is 100 La based on the EPRI guidance.[2]

4. Class 3b is conservatively categorized as LERF based on the NEI guidance and previously approved EPRI methodology.[2, §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.[2 , 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.

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

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

for evolutionary light water reactor designs.[2, §2.2]

9. The interval between ILRTs at the original licensing basis of 3 tests in 10 years is taken to be three years. This value is consistent with the EPRI guidance report. [2]

10. The likelihood of liner flaw growth over the extended period without benefit of visual inspection is estimated to double every five years.[14] This assumption is generic in nature and does not depend on any plant specific inputs and is used in the EPRI guidance. [2] As such, the doubling of the liner flaw likelihood in the Sequoyah 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] 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 ILRT). All industry events have been detected through visual inspection. Five percent visual failure detection is a conservative assumption.

12. The EPRI guidance[2] states 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.

13. The containment basemat liner is assumed to be uninspectable consistent with the EPRI methodology. [2]

14. Since a larger assumed Containment ILRT pressure yields a worse result in the corrosion sensitivity analysis, an upper bound for containment pressure during an ILRT is used in the analysis. Based on reference[21, §6.0.6] the ILRT containment pressure range is 12.0 -

12.7 psig, which corresponds to 26.7 - 27.4 psia. Accordingly, the upper bound pressure selected will be taken to be slightly larger than the 12.7 ps ig value. This value is considered reasonable because the test range is limited by procedure.

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

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

1.200 and is relevant to the ILRT 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 submittals.[28]

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2. The Sequoyah Level 1 a nd Level 2 i nternal events PRA models provide representative results.

3. Dose results for the containment failures modeled in the PRA can be characterized by information provided in NUREG/CR-4551.[9 ,10] Since Sequoyah was the representative plant in the NUREG the population doses are judged to be applicable to this analysis.

4. Accident classes describing radionuclide release end-states are defined consistent with the EPRI methodology[2] and are summarized in section 6 of this calculation.

5.0 Methodology The methodology employed is in accordance with NEI 94-01, Revision 3-A[5] 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[12] This methodology is similar to that presented in the EPRI guidance[1] as specified in NEI 94-01.[5]

A simplified bounding analysis approach is used in the methodology to evaluate the risk impact on increasing the ILRT Type A interval from the current licensing basis of one test in ten years to the proposed licensing basis of one test in fifteen 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 1 = FREQINTACT - FREQClass 3a - FREQClass 3b where; Class 3a = small containment liner leakage Class 3b = large containment liner leakage

• Accident progression sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B 2 or Type C 3 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 4 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) 1 The adjustment to Class 1 is necessary to maintain the sum of the frequencies equal to CDF.

2 Type B tests measure component leakage across pressure retaining boundaries, e.g., gaskets, expansion bellows and air locks.

3 Type C tests measure component leakage rates across containment isolation valves.

4 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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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval

- 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 1 presents detailed information regarding the EPRI accident classes[2, §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 Page E2-18 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Table 1 Detailed Description of EPRI Accident Classes Population Dose-EPRI Population Dose Description[2, §4.3] Frequency Leakage Rate (person-Class (person-rem) 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 damage sequences where containment Calculated Value isolation is successful and long-term containment heat removal Value from 1 capability is available. The frequency of an intact containment is La NUREG/CR-4551 DOSEClass 1

• FClass 1 FClass 1 =

established on the individual plants PRA. For Class 1 sequences, it is CDFIntact - F3a - F3b 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 From Plant PRA accident progression bins for which a pre-existing leakage due to failure From Value from 2 to isolate the containment occurs. These sequences are dominated by FClass 2 =

Plant PRA NUREG/CR-4551 DOSEClass 2

• FClass 2 failure-to-close of large (>2 diameter) containment isolation valves. PLargeCI

• CDFTotal SMALL PRE-EXISTING LEAK IN CONTAINMENT - all core Calculated Value damage accident progression bins with a pre-existing leakage in the FClass 3a = (Class 1 dose for 3a containment structure in excess of normal leakage. Small leaks are 10 La La)

• 10 DOSE3a

• F3a PClass 3a

• CDF 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 in the FClass 3b = (Class 1 dose for 3b containment structure in excess of normal leakage. Large leaks are 100 La La)

• 100 DOSE3b

• F3b PClass 3b* CDF 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 4 these failures are detected by Type B tests and their frequency is very N/A N/A N/A N/A 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 occurs. Because 5 these failures are detected by Type C tests and their frequency is very N/A N/A N/A N/A low compared with the other classes, this group is not evaluated further.

The frequency of Class 5 sequences is subsumed into Class 7, where it contributes insignificantly.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Population Dose-EPRI Population Dose Description[2, §4.3] Frequency Leakage Rate (person-Class (person-rem) rem/rx-yr)

CONTAINMENT ISOLATION FAILURES (DEPENDENT FAILRUES 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 to isolate the 6 containment occurs. These sequences are dominated by misalignment of N/A N/A N/A N/A containment isolation valves following test/maintenance evolutions. i.e.,

human error. All other failure modes are bounded by the Class 2 assumption.

SEVERE ACCIDENT PHENMENA - INDUCED FAILURES - all core From Plant PRA damage accident progression bins in which containment failure induced FClass 7 = From Value from 7 by severe accident phenomena occurs (e.g., hydrogen combustion and Plant PRA NUREG/CR-4551 DOSE7

• F7 CDFCFL + CDFCFE direct containment heating).

CONTAINMENT BYPASS - all core damage accident progression bins From Plant PRA in which containment bypass occurs. Each plants PRA is used to FClass 8 = From Value from 8 determine the containment bypass contribution. Contributors include Plant PRA NUREG/CR-4551 DOSE8

• F8 CDFISLOCA+ CDFSGTR ISLOCA and SGTR (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 Page E2-20 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval The risk metrics used to evaluate the impact of a proposed change on pl ant 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]),

• and, the change in the conditional containment failure probability (CCFP).

The acceptance criteria:

• LERF, RG 1.174[12]

• Population Dose<1.0 person-rem or <0.1% Increase - whichever one is less restrictive[4, §2.2]

• CCFP 1.5% (Assumption 8)

Additionally, the EPRI guidance also lists the change in core damage frequency as a measure to be considered; however, 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 ILRT test interval.

The overall methodology[2 , §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[2, §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 circumstances (i.e., ILRT Type B and Type C tests) and types of failures such as simultaneous failure of redundant isolation valves are not impacted by c hanges in the ILRT Type A frequency.[2] The baseline risk is determined as follows:

• The plant-specific Sequoyah Level 2 PRA Release categories[17] 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 Sequoyah L2 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 Page E2-21 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval extending the test interval. The increase in leakage contribution is subtracted to obtain the expected no containment failure outcome frequency as follows:

Class 1 Frequency = CDFINTACT - Class 3a Frequency - Class 3b Frequency 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.[2, §4.3] The frequencies for Class 3a and Class 3b are determined as follows:

Class 3a Frequency = CDF

• Class 3a leakage probability Class 3b Frequency = CDF

• Class 3b leakage probability 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.[2, §3.5]

According to the EPRI guidance[2] 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 large Type A containment 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 be 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 sequence 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.[2, §4.2.1]

Core damage accident progression end-states are developed for the Sequoyah PRA Level 2 results.[17] 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 Page E2-22 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval containment isolation failures prior to core damage combined with effective mitigation of fission product releases. there is no containment isolation failures prior to core damage.

5.2 Step 2 - Develop the Baseline Population Dose Per Year In step 2[2, §4.2.2] the baseline dose/yr corresponding to the current licensing basis ILRT testing interval (1-in-10 years) is estimated. Sequoyah specific estimates of population dose were developed in support of Severe Accident Mitigation Alternatives (SAMAs) for License Renewal based on 2000 Census data. After that calculation was created, the 2010 Census data was published. The calculated data was verified against the 2010 Census (permanent population) by comparing the multiplier of the calculated data (not including transient population) with the permanent population Census data for the counties within the 50-mile radius of Sequoyah. The verification showed that the calculated data based on 2000 Census results was accurate to within 0.05% of the 2010 C ensus data. Further, the calculated data was conservative, by slightly overestimating the actual population. [18, §7.3.2] Based on these results, the calculated data is acceptable for use in this calculation.

SECPOP2000 was used to determine the population within 50 miles of Sequoyah Nuclear Plant as an input to WinMACCS. SECPOP2000 is an Environmental Protection Agencys (EPA) computer program that has been used to calculate population estimates since 1973.

SECPOP2000 performs site analysis to evaluate population, land use, and economic data on a polar grid centered on a prescribed site (i.e., Sequoyah Nuclear Plant). NUREG/CR-6525 provides verification of SECPOP2000 codes by comparing them with licensee provided population data. SECPOP2000 also agrees well with census estimates from other sources.

However, SECPOP2000 does not account for transient populations. In order to account for transients, transient population estimates from TVASEQ005-CALC-001 Rev. 0 w ere used.[18,

§7.3.2]

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.[2,§4.2.2]

1. From the Sequoyah specific Level 2 results,[17] 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),[20] 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 IPE 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 Page E2-23 of 77

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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,[2, §4.2.3] the risk impact associated with the change in ILRT intervals is described.

1. Determine the change in probability of leakage detectable only by ILRT (Classes 3a and 3b) for the new surveillance intervals of interest. NUREG 1493[7] states that relaxing the ILRT frequency from three in ten years to one in ten years will increase the average time that a leak that is detectable only by ILRT goes undetected from 18 to 60 months (1/2 the surveillance interval), 60/18 = 3.33 fold increase. Therefore, relaxing the ILRT testing frequency from three in ten years to one in fifteen years will increase the average time that a leak that is detectable only by ILRT goes undetected 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), 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,[2, 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 is CCFP is described.

5.4.1 Evaluate Risk Impact - Change in LERF The risk associated with extending the ILRT interval involves a p otential 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 existing 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[12] for LERF acceptance guidelines.

LERF = (frequency class 3b interval x) - (frequency class 3b baseline)

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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,[2, §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[14] investigates how an age-related degradation mechanism can be factored into the risk impact associated with longer ILRT 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,[14] occurrences of through wall liner corrosion related defects had been found between September 1996 i mplementation 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.

5.5.1 Containment Overpressure The Sequoyah 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 ILRT 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 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[2,§5] provides an example of the technical approach for the assessment of external events LERF.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 6.0 Inputs In this section inputs from the Sequoyah Level 2 PRA are provided and relationship to the corresponding EPRI accident class is given.

Table 2 presents the EPRI release classifications and the interpretation for assignment to the Sequoyah Release categories.

Table 3 presents the Sequoyah Release categories, descriptions and mapping to the corresponding EPRI accident class.

Section 6.1 presents the decomposition of the Sequoyah accident sequences and EPRI classification.

Table 4 presents the decomposition of the Sequoyah accident sequences, frequencies and corresponding EPRI classification.

Table 5 presents the EPRI accident class frequencies, the Total CDF and the Baseline CDF (which excludes Class 6).

To determine how to Sequoyah release categories relate to the eight EPRI accident classifications the definitions are interpreted and documented in Table 2.

Table 2 EPRI Release Classes (Containment Failure Classifications)

EPRI Interpretation for Assigning Description Class Sequoyah Release Category Containment remains intact with containment 1 Intact containment bins initially isolated Isolation faults that are related to a loss of power Dependent failure modes, or common cause 2 or other isolation failure mode that is not a direct failures failure of an isolation component Independent containment isolation failures due to Isolation failures identified by Type A testing, 3

Type A related failures Large or Small 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 Isolation failure with scrubbing or small isolation 6 Other penetration failures fails Early and Late containment failure sequences as a 7 Induced by severe accident phenomena result of hydrogen detonation or other early phenomena 8 Bypass Bypass sequences, ISLOCA or SGTR Page E2-26 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval The Sequoyah Level 2 accident sequences are parsed into seven release categories that represent the summation of individual accident categories due to similar characteristics. Table 3 presents the seven release categories, descriptions[16] and mapping to the corresponding EPRI class.[2, §4.3]

Table 3 Sequoyah 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 7 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 Sequoyah accident release categories into the individual Level 2 accident sequences, their descriptions, corresponding EPRI class and the frequency is provided in Table 4.

Table 4 Decomposition of Sequoyah LERF Frequency and EPRI Classification L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class Non-station blackout scenario. Failure to depressurize the RCS resulting in a BLERF-001 5.25E-10 1.82E-10 8 thermally-induced steam generator tube rupture and a large, early release.

BLERF-002 N/A <1.00E-11 <1.00E-11 N/A 5 Non-station blackout scenario.

BLERF-003 1.70E-07 1.78E-07 8 Containment experiences a large bypass.

Station blackout scenario. Failure to depressurize the RCS early, a thermally-BLERF-004 3.85E-07 3.82E-07 8 induced steam generator tube rupture occurs resulting in a large, early release.

5 Sequences that have a result of 0.0 had a branch in the accident progression of 0.0. Those that are <1.0E-11 had results that were less than the truncation level used, therefore are given as, i.e., <1.0E-11 and have negligible or no influence on the results. Therefore, N/A or not-applicable is used for the EPRI classification and the description as there is no value added to the analysis to populate that information.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class Station blackout scenario. A pressure-BLERF-005 induced steam generator tube rupture 3.15E-08 3.28E-08 8 occurs resulting in a large, early release.

Non-station blackout scenario. There is no containment bypass or containment isolation failure. AFW is successful.

HLERF-001 RCS pressure is high and late RCS 6.83E-08 6.92E-08 7 depressurization fails. ARFs are successful. Results in early containment failure.

HLERF-002 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. RCS depressurization fails late; however, HLERF-003 containment air return fans are successful. 8.93E-09 9.28E-09 7 The hydrogen igniters fail; however, there is no detonation. Containment fails early.

HLERF-004 N/A 0.00E+00 0.00E+00 N/A HLERF-005 N/A <1.00E-11 <1.00E-11 N/A HLERF-006 N/A <1.00E-11 <1.00E-11 N/A HLERF-007 N/A 0.00E+00 0.00E+00 N/A HLERF-008 N/A <1.00E-11 <1.00E-11 N/A HLERF-009 N/A 0.00E+00 0.00E+00 N/A HLERF-010 N/A <1.00E-11 <1.00E-11 N/A HLERF-011 N/A <1.00E-11 <1.00E-11 N/A HLERF-012 N/A 0.00E+00 0.00E+00 N/A HLERF-013 N/A <1.00E-11 <1.00E-11 N/A HLERF-014 N/A 0.00E+00 0.00E+00 N/A HLERF-015 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. No SGTRs; however, RCS Depressurization fails as well as the containment air HLERF-016 recirculation fans. Hydrogen igniters are 1.75E-10 <1.00E-11 7 available and there is no direct containment heating; however, containment fails early.

HLERF-017 N/A 0.00E+00 0.00E+00 N/A HLERF-018 N/A <1.00E-11 <1.00E-11 N/A HLERF-019 N/A 0.00E+00 0.00E+00 N/A HLERF-020 N/A <1.00E-11 <1.00E-11 N/A Page E2-28 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class HLERF-028 N/A <1.00E-11 <1.00E-11 N/A HLERF-029 N/A <1.00E-11 <1.00E-11 N/A HLERF-030 N/A <1.00E-11 <1.00E-11 N/A Station blackout scenario. RCS depressurization fails resulting in a high pressure melt ejection event. The HLERF-038 hydrogen igniters are not available ; 3.68E-07 3.82E-07 7 however, there is no hydrogen detonation.

containment failure occurs early resulting in a large, early release.

Station blackout scenario. RCS depressurization fails resulting in a high pressure melt ejection event. Hydrogen HLERF-039 igniters are not available; however, there 5.78E-07 6.01E-07 7 is no hydrogen detonation. Direct containment heating occurs resulting in a large, early release.

Station blackout scenario. RCS depressurization fails resulting in a high HLERF-040 pressure melt ejection event. A hydrogen 2.10E-09 2.18E-09 7 detonation occurs resulting in a large, early release.

Non-station blackout scenario. There is no containment bypass. There is a large ILERF-001 containment isolation failure of greater 9.28E-08 9.10E-08 2 than 2 inches resulting in a large, early release.

Non-station blackout scenario. RCS depressurization fails late; however, containment air return fans are successful.

INTACT-001 1.12E-06 1.15E-06 1 The hydrogen igniters are available and containment heat removal is successful.

Containment remains intact.

Non-station blackout scenario. RCS depressurization fails late; however, containment air return fans are successful.

INTACT-002 The hydrogen igniters fail; however, there 3.50E-08 3.64E-08 1 is no detonation. Containment heat removal is successful and containment remains intact.

Non-station blackout scenario. RCS depressurization fails late and the INTACT-003 containment air return fans fail. Hydrogen 1.75E-10 1.82E-10 1 igniters and containment heat removal is successful. Containment remains intact.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class INTACT-004 N/A 0.00E+00 <1.00E-11 N/A Non-station blackout scenario. Late RCS depressurization and the containment air return fans are successful. Hydrogen INTACT-005 3.93E-06 4.00E-06 1 igniters are available and containment heat removal is successful. Containment remains intact.

Non-station blackout scenario. Late RCS depressurization and the containment air return fans are successful. Hydrogen INTACT-006 igniters are unavailable; however, 0.00E+00 1.82E-10 1 detonation does not occur. Containment heat removal is successful. Containment remains intact.

INTACT-007 N/A 0.00E+00 <1.00E-11 N/A INTACT-008 N/A 0.00E+00 <1.00E-11 N/A INTACT-009 N/A 0.00E+00 <1.00E-11 N/A INTACT-010 N/A 0.00E+00 <1.00E-11 N/A Non-station blackout scenario. No SGTRs; however, RCS Depressurization fails as well as the containment air INTACT-011 recirculation fans. Hydrogen igniters are 1.75E-10 1.46E-09 1 available, containment heat removal is successful, therefore containment remains intact.

INTACT-012 N/A 0.00E+00 <1.00E-11 N/A INTACT-013 N/A 0.00E+00 <1.00E-11 N/A INTACT-014 N/A 0.00E+00 <1.00E-11 N/A INTACT-015 N/A 0.00E+00 <1.00E-11 N/A INTACT-016 N/A 0.00E+00 <1.00E-11 N/A Non-station blackout scenario. Early RCS depressurization and containment air return fans are successful. Hydrogen INTACT-017 1.14E-06 1.38E-06 1 igniters are available and containment heat removal is successful; therefore, containment remains intact.

Non-station blackout scenario. Early RCS depressurization and containment air INTACT-018 return fans are successful. Hydrogen <1.00E-11 1.82E-10 1 igniters fail; however, no detonation occurs. Containment heat removal is successful therefore containment remains Page E2-30 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class intact.

Non-station blackout scenario. Early RCS depressurization is successful; however, the containment air return fans fail. The INTACT-019 7.00E-10 9.46E-09 1 hydrogen igniters are successful as well as the containment heat removal function; therefore, containment remains intact.

INTACT-020 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. Core damage has occurred, the RCS is at low pressure, the containment air return fans INTACT-021 are successful and the hydrogen igniters 8.58E-08 8.92E-08 1 are available. Containment heat removal is successful; therefore, containment remains intact.

INTACT-022 N/A <1.00E-11 <1.00E-11 N/A INTACT-023 N/A <1.00E-11 <1.00E-11 N/A INTACT-024 N/A <1.00E-11 <1.00E-11 N/A INTACT-028 N/A 0.00E+00 0.00E+00 N/A INTACT-032 N/A <1.00E-11 <1.00E-11 N/A INTACT-036 N/A 0.00E+00 0.00E+00 N/A INTACT-040 N/A <1.00E-11 <1.00E-11 N/A INTACT-044 N/A 0.00E+00 0.00E+00 N/A INTACT-048 N/A <1.00E-11 <1.00E-11 N/A INTACT-052 N/A 0.00E+00 0.00E+00 N/A INTACT-056 N/A <1.00E-11 <1.00E-11 N/A INTACT-060 N/A 0.00E+00 0.00E+00 N/A INTACT-064 N/A 0.00E+00 0.00E+00 N/A INTACT-068 N/A 0.00E+00 0.00E+00 N/A INTACT-072 N/A <1.00E-11 <1.00E-11 N/A LATE-001 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. RCS depressurization fails late; however, containment air return fans are successful.

LATE-002 8.75E-10 1.09E-09 7 The hydrogen igniters are available and containment heat removal fails resulting in a late release.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class LATE-003 N/A 0.00E+00 0.00E+00 N/A LATE-004 N/A <1.00E-11 <1.00E-11 N/A LATE-005 N/A 0.00E+00 0.00E+00 N/A LATE-006 N/A <1.00E-11 <1.00E-11 N/A LATE-007 N/A 0.00E+00 0.00E+00 N/A LATE-008 N/A <1.00E-11 <1.00E-11 N/A LATE-009 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Late RCS depressurization and the containment air return fans are successful. Hydrogen LATE-010 5.60E-09 5.64E-09 7 igniters are available; however, containment heat removal fails resulting in a late release.

LATE-011 N/A 0.00E+00 0.00E+00 N/A LATE-012 N/A <1.00E-11 <1.00E-11 N/A LATE-013 N/A 0.00E+00 0.00E+00 N/A LATE-014 N/A <1.00E-11 <1.00E-11 N/A LATE-015 N/A 0.00E+00 0.00E+00 N/A LATE-016 N/A <1.00E-11 <1.00E-11 N/A LATE-017 N/A 0.00E+00 0.00E+00 N/A LATE-018 N/A <1.00E-11 <1.00E-11 N/A LATE-019 N/A 0.00E+00 0.00E+00 N/A LATE-020 N/A <1.00E-11 <1.00E-11 N/A LATE-021 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. No SGTR; however, RCS Depressurization fails as well as the containment air recirculation LATE-022 4.20E-09 2.00E-09 7 fans. Hydrogen igniters are available, containment heat removal fails resulting in a late release.

LATE-023 N/A 0.00E+00 0.00E+00 N/A LATE-024 N/A <1.00E-11 <1.00E-11 N/A LATE-025 N/A 0.00E+00 0.00E+00 N/A LATE-026 N/A <1.00E-11 <1.00E-11 N/A LATE-027 N/A 0.00E+00 0.00E+00 N/A Page E2-32 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class LATE-028 N/A <1.00E-11 <1.00E-11 N/A LATE-029 N/A 0.00E+00 0.00E+00 N/A LATE-030 N/A <1.00E-11 <1.00E-11 N/A LATE-031 N/A 0.00E+00 0.00E+00 N/A LATE-032 N/A <1.00E-11 <1.00E-11 N/A LATE-033 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Early RCS depressurization and containment air return fans are successful. Hydrogen LATE-034 3.15E-07 4.91E-07 7 igniters are available: however, containment heat removal fails resulting in a late release.

LATE-035 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Early RCS depressurization and containment air return fans are successful. Hydrogen LATE-036 1.75E-10 1.82E-10 7 igniters fail; however, no detonation occurs. Containment heat removal fails resulting in a late release.

LATE-037 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Early RCS depressurization is successful. The containment air return fans fails; LATE-038 however, the hydrogen igniters are 2.10E-08 1.40E-08 7 available. Containment heat removal is successful resulting in an intact containment.

LATE-039 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Early RCS depressurization is successful and containment air return fans are unavailable. Hydrogen igniters are not LATE-040 1.75E-10 <1.00E-11 7 available; however, hydrogen detonation does not occur. Containment does not fail early. Containment heat removal fails resulting in a late release.

Non-station blackout scenario. Core damage has occurred, the RCS is at low pressure, the containment air return fans LATE-041 are successful and the hydrogen igniters 0.00E+00 0.00E+00 7 are available. Containment heat removal is successful; however, basemat melt-through occurs resulting in a late release.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class Non-station blackout scenario. Core damage has occurred, the RCS is at low pressure, the containment air return fans LATE-042 1.93E-06 2.18E-06 7 are successful and the hydrogen igniters are available. Containment heat removal fails which results in a late release.

LATE-043 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Core damage has occurred and the RCS is at low pressure, the hydrogen igniters are LATE-044 2.98E-09 4.73E-09 7 not available. Detonation does not occur; however, containment heat removal fails resulting in a late release.

LATE-045 N/A 0.00E+00 0.00E+00 N/A Non-station blackout scenario. Core damage has occurred and the RCS is at low pressure. The containment air return LATE-046 1.58E-09 4.00E-08 7 fans fail; however, the hydrogen igniters are available. Containment heat removal fails resulting in a late release.

LATE-047 N/A 0.00E+00 0.00E+00 N/A LATE-048 N/A <1.00E-11 <1.00E-11 N/A LATE-055 N/A 0.00E+00 0.00E+00 N/A LATE-056 N/A 0.00E+00 0.00E+00 N/A LATE-063 N/A 0.00E+00 0.00E+00 N/A LATE-064 N/A <1.00E-11 <1.00E-11 N/A LATE-071 N/A 0.00E+00 0.00E+00 N/A LATE-072 N/A 0.00E+00 0.00E+00 N/A LATE-079 N/A 0.00E+00 0.00E+00 N/A LATE-080 N/A <1.00E-11 <1.00E-11 N/A LATE-087 N/A 0.00E+00 0.00E+00 N/A LATE-088 N/A 0.00E+00 0.00E+00 N/A LATE-095 N/A 0.00E+00 0.00E+00 N/A Station blackout scenario. RCS depressurization fails resulting in a high pressure melt ejection event. The LATE-096 hydrogen igniters are not available; 1.30E-06 1.33E-06 7 however, there is no hydrogen detonation.

Containment heat removal fails resulting in a late release.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class LATE-103 N/A 0.00E+00 0.00E+00 N/A LATE-104 N/A 0.00E+00 0.00E+00 N/A LATE-111 N/A 0.00E+00 0.00E+00 N/A Station blackout scenario. Early RCS depressurization fails; however, the steam generators remain intact. Late RCS depressurization is successful; however, LATE-112 not in time to prevent a low pressure melt 5.60E-06 5.82E-06 7 ejection event. The hydrogen igniters are not available; however, there is no hydrogen detonation. Containment heat removal fails resulting in a late release.

LATE-119 N/A 0.00E+00 0.00E+00 N/A LATE-120 N/A 0.00E+00 0.00E+00 N/A LATE-127 N/A 0.00E+00 0.00E+00 N/A LATE-128 N/A 0.00E+00 0.00E+00 N/A LATE-135 N/A 0.00E+00 0.00E+00 N/A LATE-136 N/A 0.00E+00 0.00E+00 N/A LATE-143 N/A 0.00E+00 0.00E+00 N/A Station blackout scenario. Core damage has occurred and the RCS is at low pressure. The hydrogen igniters are not LATE-144 3.15E-07 8.37E-08 7 available; however, no detonation occurs.

Containment heat removal fails resulting in a late release.

LLERF-001 N/A <1.00E-11 <1.00E-11 N/A LLERF-002 N/A <1.00E-11 <1.00E-11 N/A LLERF-003 N/A <1.00E-11 <1.00E-11 N/A LLERF-004 N/A <1.00E-11 <1.00E-11 N/A LLERF-005 N/A <1.00E-11 <1.00E-11 N/A LLERF-006 N/A <1.00E-11 <1.00E-11 N/A LLERF-007 N/A <1.00E-11 <1.00E-11 N/A LLERF-008 N/A <1.00E-11 <1.00E-11 N/A LLERF-009 N/A <1.00E-11 <1.00E-11 N/A LLERF-010 N/A <1.00E-11 <1.00E-11 N/A LLERF-011 N/A <1.00E-11 <1.00E-11 N/A LLERF-012 N/A <1.00E-11 <1.00E-11 N/A Page E2-35 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class LLERF-013 N/A <1.00E-11 <1.00E-11 N/A LLERF-014 N/A <1.00E-11 <1.00E-11 N/A LLERF-015 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. Early RCS depressurization is successful; the containment air return fans and the LLERF-016 1.75E-10 <1.00E-11 7 hydrogen igniters are available; however, containment fails early resulting in a large, early release.

LLERF-017 N/A <1.00E-11 <1.00E-11 N/A LLERF-018 N/A <1.00E-11 <1.00E-11 N/A LLERF-019 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. Core damage has occurred and the RCS is at low pressure, the hydrogen igniters are LLERF-020 5.25E-10 9.10E-10 7 not available. Detonation does not occur; however, containment fails resulting in a large, early release.

LLERF-021 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario. Core damage has occurred and the RCS is at low pressure. The containment air return LLERF-022 fans fail; however, the hydrogen igniters <1.00E-11 1.82E-10 7 are available. Containment fails resulting in a large, early release.

LLERF-023 N/A <1.00E-11 <1.00E-11 N/A LLERF-024 N/A <1.00E-11 <1.00E-11 N/A LLERF-029 N/A <1.00E-11 <1.00E-11 N/A LLERF-030 N/A <1.00E-11 <1.00E-11 N/A Station blackout scenario. Early RCS depressurization fails; however, the steam generators remain intact. Late RCS depressurization is successful; however, LLERF-035 not in time to prevent a low pressure melt 1.26E-06 1.29E-06 7 ejection event. The hydrogen igniters are not available: however no detonation occurs. Containment fails resulting in a large, early release.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval L2 Accident Unit 1 Unit 2 EPRI Accident Sequence Description Sequence Frequency/yr Frequency/yr Class Station blackout scenario. Early RCS depressurization fails; however, the steam generators remain intact. Late RCS LLERF-036 depressurization is successful; however, 7.35E-09 7.64E-09 7 not in time to prevent a low pressure melt ejection event. A hydrogen detonation occurs resulting in a large, early release.

LLERF-041 N/A 0.00E+00 0.00E+00 N/A LLERF-042 N/A 0.00E+00 0.00E+00 N/A LLERF-047 N/A <1.00E-11 <1.00E-11 N/A LLERF-048 N/A <1.00E-11 <1.00E-11 N/A Non-station blackout scenario.

SERF-001 Containment Isolation (Small) Failure. 1.93E-06 2.00E-06 6 No By-pass SERF-002 N/A <1.00E-11 <1.00E-11 N/A SERF-007 N/A 0.00E+00 0.00E+00 N/A SERF-008 N/A 0.00E+00 0.00E+00 N/A SERF-013 N/A 0.00E+00 0.00E+00 N/A SERF-014 N/A 0.00E+00 0.00E+00 N/A SERF-018 N/A 0.00E+00 0.00E+00 N/A SERF-019 N/A 0.00E+00 0.00E+00 N/A SERF-025 N/A 0.00E+00 0.00E+00 N/A SERF-026 N/A 0.00E+00 0.00E+00 N/A SERF-031 N/A 0.00E+00 0.00E+00 N/A SERF-032 N/A 0.00E+00 0.00E+00 N/A SERF-037 N/A 0.00E+00 0.00E+00 N/A SERF-038 N/A 0.00E+00 0.00E+00 N/A Page E2-37 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Table 5 presents the EPRI accident class frequencies based on the data from Table 4.

Table 5 EPRI Accident Class Frequencies EPRI Accident Class Totals[17]

Frequency Classification Unit 1 Unit 2 Sum of EPRI Class 1 6.30E-06 6.67E-06 Sum of EPRI Class 2 9.29E-08 9.10E-08 Sum of EPRI Class 6 1.93E-06 2.00E-06 Sum of EPRI Class 7 1.18E-05 1.23E-05 Sum of EPRI Class 8 5.88E-07 5.94E-07 Sum of LERF Sequences 2.98E-06 3.05E-06 Total Level 2 End-States (Total CDF) Including Class 6 2.07E-05 2.17E-05 Total Level 2 End-States (Baseline CDF) Excluding Class 6 1.88E-05 1.97E-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 (three-year ILRT 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 (three-year ILRT frequency) risk in terms of frequency per reactor year for the EPRI accident classes of interest.[2,

§4.2]

Section 7.1.1 presents the calculation for the frequency of Class 2 failures which consists of large containment isolation failures.

Section 7.1.2 presents Class 7 failures which consist of early and late severe accident phenomena.

Section 7.1.3 presents 8 failures 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.

Page E2-38 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 4 as Class 2 taken from the Sequoyah specific Level 2 analysis, and summed in Table 5.

Equation 1 Calculation of the Class 2 Frequency Unit 1 FREQClass_2 = Class_2 Accident Sequences

= 9.29E-08/yr Unit 2 FREQClass_2 = Class_2 Accident Sequences

= 9.10E-08/yr 7.1.2 Class 7 - Severe Accident Phenomena Class 7 represents early and late containment failure sequences involving 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 4 and summed in Table 5.

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

= 1.18E-05/yr Unit 2 FREQClass_7 = Class_7 Accident Sequences

= 1.23E-05/yr Note: Class 7 events contribute to both early and late releases and is distributed as follows:

Unit 1 FREQClass_7_LERF = 2.30E-06/yr FREQClass_7_LATE = 9.50E-06/yr Unit 2 FREQClass_7_LERF = 2.36E-06/yr FREQClass_7_LATE = 9.98E-06/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 4 and summed in Table 5. 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

= 5.88E-07/yr Unit 2 FREQClass_8 = Class_8 Accident Sequences

= 5.94E-07/yr Page E2-39 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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[2] 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 ILRT. 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[2, §4.3] 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 three tests every ten years, opposed to Sequoyahs current once per ten years frequency. The probability must be adjusted to reflect this difference which is performed later in this calculation.

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

Equation 5 Calculation of the Class 3a Failure Frequency Unit 1 FREQClass_3a = PClass_3a x CDFBaseline

= 0.0092 x 1.88E-05

= 1.73E-07/yr Unit 2 FREQClass_3a = PClass_3a x CDFBaseline

= 0.0092 x 1.97E-05/yr

= 1.82E-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 [2, §2.3]

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.

Page E2-40 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 Equation 7 Calculation of the Class 3b Failure Frequency Unit 1 FREQClass_3b = PClass_3b x CDFBaseline

= 0.0023 x 1.88E-05/yr

= 4.31E-08/yr Unit 2 FREQClass_3b = PClass_3b x CDFBaseline

= 0.0023 x 1.97E-05/yr

= 4.52E-08/yr 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)

= 1.88E-05/yr - (9.29E-08/yr + 1.18E-05/yr + 5.88E-07/yr)

= 6.30E-06/yr Unit 2 FREQClass_1 = CDFBaseline - ( FREQClass 2 + FREQClass 7 + FREQClass 8)

= 1.97E-05/yr - (9.10E-08/yr + 1.23E-05/yr + 5.94E-07/yr)

= 6.67E-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)

= 6.30E-06/yr - (1.73E-07/yr + 4.31E-08/yr)

= 6.08E-06/yr Unit 2 FREQClass 1 ADJ = FREQClass 1_BL - (FREQClass 3a_OLB + FREQClass 3b_OLB)

= 6.67E-06/yr - (1.82E-07/yr + 4.52E-08/yr)

= 6.44E-06/yr Page E2-41 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 41

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 Sequoyah plant. The following sub-steps and tables are provided in this section.

Section 7.2.1 presents the Sequoyah 50-mile radius population density Table 6 presents the estimated 2014 population density in the 50-mile radius surrounding Sequoyah Section 7.2.2 presents the NUREG/CR-4551 off-site consequence (person-rem) estimates Table 7 presents the estimates person-rem for Sequoyah source term groups reported from NUREG/CR-4551.[9] This table also assigns each source term group to an EPRI class Table 8 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 7.2.3 presents the Sequoyah off-site consequence (person-rem) estimates Table 9 presents the Unit-1 Baseline dose calculation without the breakout of Class 3 which is included in the Class 1 values.

Table 10 presents the Unit-2 Baseline dose calculation without the breakout of Class 3 which is included in the Class 1 values.

Table 11 and Table 12 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 11 presents the Unit-1 Baseline dose calculation with values for Class 3a and 3a which were taken from the Class 1 values to maintain the total CDF.

Table 12 presents the Unit-2 Baseline dose calculation with values for Class 3a and 3a which were taken from the Class 1 values to maintain CDF.

7.2.1 50-Mile Radius Population Density The estimated population within a 50-mile radius of the Sequoyah Plant was taken from calculation TVASEQ008-CALC-002[18] which was developed for plant life extension analysis for Severe Accident Mitigation Alternatives (SAMA). The basis for the population projections are described fully in that calculation. The results, in part are provided in Table 6.

Table 6 50-Mile Radius Population Density Permanent Population Transient Population Total Population

[18, Table 7.2]

2014 Data 1,128,529 96,394 1,224,924 In order to determine the dose (person-rem) for a given accident progression bin it is necessary to associate each release category with an associated release of radionuclides. The EPRI guidance

[2]

on leak-rate testing indicates that a surrogate can be applied and is acceptable in estimating risk. The guidance suggests one surrogate source is the results contained in NUREG-1150[6]

which presents results for both BWRs and PWRs. The BWR results are not considered in this Page E2-42 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 42

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval analysis since the core melt mechanics and design are substantially different as c ompared to Sequoyah.

7.2.2 NUREG/CR-4551 Off-Site Consequence (Person-Rem Estimates)

NUREG/CR-4551 Vol. 5 Rev 1 Part 1[9] provides the Level 2 analysis and off-site consequence assessment for Sequoyah. Table 4.3-1 in that report provides a summary of consequence results that includes population dose within 50-miles for internal events. The dose estimates for a 50-mile radius from the Sequoyah site are re-produced in Table 7 by the source term grouping for each outcome. These values summarize the data from NUREG/CR-4551 Vol. 5 R ev. 1 P art 1, Table 4.3-1. The NUREG values were reported in person-Sieverts and were converted to person-rem in Table 7 by multiplying Sieverts x 100. The dose estimates for each outcome associated with each source term group is averaged to obtain a r epresentative value for that source term group. Then the individual groups are averaged to obtain a class estimate.

Table 7 Reported Person-Rem Estimates for Sequoyah Source Term Groups Source PERSON-REM[9, Table 4.3-1] Dose (Person-Rem)

EPRI Class Term Average of Grouping Outcome 1 6 Outcome 2 Outcome 3 Assignment Outcomes SEQ-01 NA 3.19E+04 NA 3.19E+04 8 SEQ-02 1.26E+05 1.26E+05 1.06E+05 1.19E+05 7 SEQ-03 2.91E+05 3.17E+05 NA 3.04E+05 7 SEQ-04 7.71E+05 4.60E+05 6.32E+05 6.21E+05 8 SEQ-05 4.41E+05 6.42E+05 NA 5.42E+05 7 SEQ-06 3.79E+05 5.83E+05 NA 4.81E+05 7 SEQ-07 NA NA 1.49E+06 1.49E+06 8 SEQ-08 1.30E+06 1.02E+06 1.07E+06 1.13E+06 7 SEQ-09 9.32E+05 8.32E+05 NA 8.82E+05 7 SEQ-10 7.20E+05 1.08E+06 NA 9.00E+05 7 SEQ-11 2.57E+06 1.97E+06 3.37E+06 2.64E+06 7 SEQ-12 1.63E+06 1.45E+06 1.09E+06 1.39E+06 7 SEQ-13 1.21E+06 1.16E+06 NA 1.19E+06 7 6

The three outcomes represent the source term groups split into three subgroups. The subgroups are based on evacuation times for the general public relative to the commencement of the evacuation and the timing of the release. Outcome 1 - evacuation starts at least 30 minutes before the release, Outcome 2 - evacuation starts 30 minutes before the release up to 60 minutes after the release, and Outcome 3 - evacuation starts more than 60 minutes after the release begins.[9, Pg 3-55]

Page E2-43 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Source PERSON-REM[9, Table 4.3-1] Dose (Person-Rem)

EPRI Class Term Average of Grouping Outcome 1 6 Outcome 2 Outcome 3 Assignment Outcomes SEQ-14 1.13E+07 2.96E+06 8.20E+06 7.49E+06 2 SEQ-15 2.27E+06 2.10E+06 2.54E+06 2.30E+06 7 SEQ-16 1.38E+02 1.98E+03 NA 1.06E+03 7 SEQ-17 1.14E+03 8.09E+03 NA 4.62E+03 7 SEQ-18 1.06E+05 3.06E+05 NA 2.06E+05 7 SEQ-19 0.00E+00 0.00E+00 0.00E+00 NA N/A Note: SEQ-019 is not applied to an EPRI class since the outcomes listed in Table 7 are either 0.0 or NA.

In order to utilize this information it is necessary to convert the data to the form necessary for the ILRT analysis. This involves classification into one of three EPRI classes 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 in the following table.

Table 8 Calculation of the Sequoyah Population Dose Risk at 50-Miles Collapsed NUREG/CR-4551 NUREG/CR-Collapsed Fractional NUREG/CR- 4551 Population Dose Accident Accident APB 4551 Collapsed Population Risk at 50-Miles Progression Progression Bin Contributions APB Dose at (Person-Rem/yr -

Bin (APB) 1 to Risk2 Frequency/Yr5 Mean) 4 50-Miles6 VB, Early CF I 0.041 0.492 2.79E-07 1.76E+06 (During CD)

VB, Early CF (At II 0.012 0.144 1.12E-07 1.29E+06 VB), Alpha Mode VB >200 psi, Early III 0.107 1.284 1.95E-06 6.57E+05 CF (at VB)

VB <200 psi, Early IV 0.102 1.224 1.28E-06 9.54E+05 CF (at VB)

V VB, Late CF 0.096 1.152 2.12E-06 5.43E+05 VB, BMT, Very VI 0.21 2.520 9.54E-06 2.64E+05 Late CF VII Bypass 0.376 4.512 3.12E-06 1.44E+06 VB, No CF, No VIII 0.002 0.024 1.50E-05 1.60E+03 Bypass No VB, Early CF IX 0.052 0.624 6.14E-07 1.02E+06 (During CD)

X No VB, No CF 0.002 0.024 2.07E-05 1.16E+03 Page E2-44 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 44

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Collapsed NUREG/CR-4551 NUREG/CR-Collapsed Fractional NUREG/CR- 4551 Population Dose Accident Accident APB 4551 Collapsed Population Risk at 50-Miles Progression Progression Bin Contributions APB Dose at (Person-Rem/yr -

Bin (APB) 1 to Risk2 Frequency/Yr5 Mean) 4 50-Miles6 Total From All 1.000 12.03 5.58E-05 APBs:

1. The accident progression bins used in the EPRI guidance [1 Table 5-5] were taken from Surry data, the corresponding Sequoyah ABP is used for this analysis, including Vessel Breach (VB), Alpha (Steam Explosion) and Early CF (Containment Failure).[9, Table 5.1-3]

2. The Mean Fractional Contributions to Risk (MFCR) are taken from NUREG/CR-4551 Vol 5 Rev 1 Part 1. [9, Figure 2.5-3]

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. [9 Table 5.1-1]

4. The contribution for a given APB is the product of the total PDR50 (12.0 person-rem/rx-yr) and the fractional APB contribution).

5. NUREG/CR-4551 provides the conditional probabilities of the collapsed APBs.[9, Figure 2.5 -3] These conditional probabilities (0.002 for VB, ALPHA, Early CF) are multiplied by the total internal CDF (from the NUREG),

also called 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 third column (0.144) of this table by the collapsed APB (VB, Alpha, Early CF) frequency (1.12E-07) shown in the fourth column of this table.

7.2.3 Sequoyah Specific Off-Site Consequence (Person-Rem Estimates)

EPRI guidance[2 §5.2.2] uses a multiplication factor to develop the design basis leakage value (La) based on generic information that provides comparative local population ratios. Since Sequoyah is a r eference plant in the NUREG/CR-4551,[9] an adjustment to the population dose is not necessary.

To determine the applicable population dose for Sequoyah, the population dose methodology for the Surry collapsed accident progression bins are referenced. This analysis method from the EPRI guidance [2, Table 5-5] was used to estimate the population density within a 50-mile radius of the Sequoyah plant. The corresponding inputs for Sequoyah are inserted into the equations. The results are provided in Table 9 and Table 10 which present the Sequoyah Unit 1 a nd Unit 2 release category mapping for the eight EPRI accident classes. Dose (person-rem) per year is the product of the frequency (per year) and the person-rem for an accident class.

Table 9 U1 - Baseline Dose Calculation (Without 3a & 3b)

EPRI Baseline Dose EPRI Description Person-Rem/yr Class Frequency/yr (Person-Rem) 1 Intact Containment 6.30E-06 2.76E+03 7 1.74E-02 7

The dose (person-rem) attributed to Class 1 is the sum of accident progression bins VIII and X from Table 8.

Page E2-45 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 2 Large Containment Isol Failures 9.29E-08 7.49E+06 8 6.96E-01 3a Small Isol Failures (Liner Breach) (1) (1) (1) 3b Large Isolation failures (Liner Breach) (1) (1) (1)

Small Isolation Failures - Failure-to-4 (2) (2) (2)

Seal (Type B)

Small Isolation Failures - Failure-to-5 (2) (2) (2)

Seal (Type C)

Containment Isolation Failures 6 (2) (2) (2)

(Dependent Failure, Personnel Errors)

Severe Accident Phenomena Induced 7 1.18E-05 8.63E+05 1.02E+01 Failure (Early)

Containment Bypass (ISLOCA, 8 5.88E-07 7.14E+05 4.20E-01 SGTR)

Total: 1.88E-05 1.131E+01

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 ILRT interval frequency; therefore, are not included in the analysis.

Table 10 U2 - Baseline Dose Calculation (Without 3a & 3b)

EPRI Dose EPRI Description Frequency/yr Person-Rem/yr Class (Person-Rem) 1 Intact Containment 6.67E-06 2.76E+03 1.84E-02 Large Containment Isolation 2 9.10E-08 7.49E+06 6.82E-01 Failures Small Isolation failures (Liner 3a (1) (1) (1)

Breach)

Large Isolation failures (Liner 3b (1) (1) (1)

Breach)

Small Isolation Failures -

4 (2) (2) (2)

Failure-to-Seal (Type B)

Small Isolation Failures -

5 (2) (2) (2)

Failure-to-Seal (Type C) 8 Class 2, 7 and 8 values are taken from Table 7.

Page E2-46 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 46

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Containment Isolation Failures 6 (Dependent Failure, Personnel (2) (2) (2)

Errors)

Severe Accident Phenomena 7 1.23E-05 8.63E+05 1.07E+01 Induced Failure (Early)

Containment Bypass (ISLOCA, 8 5.94E-07 7.14E+05 4.24E-01 SGTR)

Total: 1.97E-05 1.178E+01 Table 11 and Table 12 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 11 U1 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution)

EPRI Baseline Dose EPRI Description Person-Rem/yr Class Frequency/yr (Person-Rem) 1 Intact Containment(1) 6.08E-06 2.76E+03 1.68E-02 Large Containment Isolation 2 9.29E-08 7.49E+06 6.96E-01 Failures Small Isolation failures (Liner 3a 1.73E-07 2.76E+04 9 4.78E-03 Breach)

Large Isolation failures (Liner 3b 4.31E-08 2.76E+05 10 1.19E-02 Breach)

Severe Accident Phenomena 7 1.18E-05 8.63E+05 1.02E+01 Induced Failure (Early)

Containment Bypass (ISLOCA, 8 5.88E-07 7.14E+05 4.20E-01 SGTR)

Total: 1.88E-05 1.133E+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 12 U2 - Baseline (Adjusted) Dose Calculation (With 3a & 3b Contribution)

EPRI Dose EPRI Description Frequency/yr Person-Rem/yr Class (Person-Rem) 9 The dose attributed (assumed) to the 3a liner leakage is 10x the intact containment dose burden.

10 The dose attributed (assumed) to the 3b liner leakage is 100x the intact containment dose burden.

Page E2-47 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 1 Intact Containment(1) 6.44E-06 2.76E+03 1.78E-02 Large Containment Isolation 2 9.10E-08 7.49E+06 6.82E-01 Failures Small Isolation failures (Liner 3a 1.82E-07 2.76E+04 5.01E-03 Breach)

Large Isolation failures (Liner 3b 4.52E-08 2.76E+05 1.25E-02 Breach)

Severe Accident Phenomena 7 1.23E-05 8.63E+05 1.07E+01 Induced Failure (Early)

Containment Bypass (ISLOCA, 8 5.94E-07 7.14E+05 4.24E-01 SGTR)

Total: 1.97E-05 1.179E+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.

7.3 Step 3 - Risk Impact Evaluation In this step, the risk associated with the change in ILRT 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 three step process.

The current surveillance testing requirement of Type A testing and allowed by 10CFR50, Appendix J is at least once-in-10 years based on an acceptable performance history 11 and represents the current licensing basis for Sequoyah. Extending the Type A ILRT interval from 3-in-10 years (original licensing basis) to once-In-10 years 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 factor of 3.33 increase. Therefore, considering the proposed licensing basis of extending the ILRT Type A test interval from 3-In-10 years to once-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 factor of five increase.

The following sub-steps are contained within Step 3.

Section 7.3.1 presents the risk impact for a 1-in-10 years test interval Table 13 presents the Unit 1 frequency, dose-rate and dose for a testing interval of 1-in-10 years Table 14 presents the Unit 2 frequency, dose-rate and dose for a testing interval of 1-in-10 years 11 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.

Page E2-48 of 77

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Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Section 7.3.2 presents the risk impact for a 1-in-15 years test interval Table 15 presents the Unit 1 frequency, dose-rate and dose for a testing interval of 1-in-15 years Table 16 presents the Unit 2 frequency, dose-rate and dose for a testing interval of 1-in-15 years Section 7.3.3 presents the dose-rate increase and percentile increase Table 17 presents the Unit 1 Class 1 population dose rate increase due to extending the ILRT interval Table 18 presents the Unit 1 Class 3a population dose rate increase due to extending the ILRT interval Table 19 presents the Unit 1 Class 3b population dose rate increase due to extending the ILRT interval Table 20 presents the Unit 1 total population dose rate increase due to extending the ILRT interval Equation 10 calculates the percent increase in total population dose Table 21 presents the Unit 2 Class 1 population dose rate increase due to extending the ILRT interval Table 22 presents the Unit 2 Class 3a population dose rate increase due to extending the ILRT interval Table 23 presents the Unit 2 Class 3b population dose rate increase due to extending the ILRT interval Table 24 presents the Unit 2 total population dose rate increase due to extending the ILRT interval Equation 10 calculates the percent increase in total population dose 7.3.1 Risk Impact - Once-in-10 Years Test Interval Based on t he 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 is determined by multiplying the Class 3 accident frequency by a factor or 3.33. Additionally, the Class 1 f requency is adjusted to maintain the overall core damage frequency constant. The results of this analysis are presented in Table 13 and Table 14 for Units 1 and 2, respectively.

Table 13 U1 - Testing Once-in-10 Years Risk Profile EPRI Dose Description Freq/yr Person-Rem/yr Class (Person-Rem) 1 Intact Containment(1) 5.58E-06 2.76E+03 1.54E-02 2 Large Cont. Isol. Failures 9.29E-08 7.49E+06 6.96E-01 Small Isolation Failures (Liner 3a 5.76E-07 2.76E+04 1.59E-02 Breach) Assumed 10 La Page E2-49 of 77

Calculation No. MDN 000 999 2014 000162 Rev: 000 Plant: Sequoyah Page: 49

Subject:

Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Large Isolation Failures (Liner 3b 1.43E-07 2.76E+05 3.96E-02 Breach) Assumed 100 La Severe Accident Phenomena 7 1.18E-05 8.63E+05 1.02E+01 Induced Failure (Early & Late)

Containment Bypass (ISLOCA, 8 5.88E-07 7.14E+05 4.20E-01 SGTR)

Total: 1.88E-05 1.137E+01

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

Table 14 U2 - Testing Once-in-10 Years Risk Profile EPRI Dose Description Freq/yr Person-Rem/yr Class (Person-Rem) 1 Intact Containment (1) 5.92E-06 2.76E+03 1.63E-02 2 Large Cont. Isolation Failures 9.10E-08 7.49E+06 6.82E-01 Small Isolation failures (Liner 3a 6.05E-07 2.76E+04 1.67E+02 Breach) 3b Large Isolation failures (Liner 1.50E-07 2.76E+05 4.15E-02 Breach)

Severe Accident Phenomena 7 1.23E-05 8.63E+05 1.07E-01 Induced Failure (Early & Late)

Containment Bypass (ISLOCA, 8 5.94E-07 7.14E+05 4.24E-01 SGTR)

Total: 1.97E-05 1.183E+01

1. The PRA frequency of Class 1 has 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 ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18-months to 90-months); therefore the baseline data for Class 3 is multiplied by a factor of 5[2, §4.2.3] and Class 1 is adjusted to conserve CDF.

The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally, the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 15 and Table 16, for Units 1 and 2, respectively.

Table 15 U1 - Testing Once-in-15 Years Risk Profile EPRI Dose Description Freq/yr Person-Rem/yr Class (Person-Rem) 1 Intact Containment(1) 5.22E-06 2.76E+03 1.44E-02 Page E2-50 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval EPRI Dose Description Freq/yr Person-Rem/yr Class (Person-Rem)

Large Containment 2 9.29E-08 7.49E+06 6.96E-01 Isolation Failures Small Isolation Failures 3a 8.65E-07 2.76E+04 2.39E-02 (Liner Breach)

Large Isolation Failures 3b 2.15E-07 2.76E+05 5.94E-02 (Liner Breach)

Severe Accident 7 Phenomena Induced 1.18E-05 8.63E+05 1.02E+01 Failure (Early & Late)

Containment Bypass 8 5.88E-07 7.14E+05 4.20E-01 (ISLOCA, SGTR)

Total: 1.88E-05 1.139E+01

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

Table 16 U2 - Testing Once-in-15 Years Risk Profile EPRI Dose Description Freq/yr Person-Rem/yr Class (Person-Rem) 1 Intact Containment(1) 5.54E-06 2.76E+03 1.53E-02 Large Containment Isolation 2 9.10E-08 7.49E+06 6.82E-01 Failures Small Isolation Failures 3a 9.08E-07 2.76E+04 2.51E-02 (Liner Breach)

Large Isolation Failures 3b 2.26E-07 2.76E+05 6.24E-02 (Liner Breach)

Severe Accident Phenomena 7 Induced Failure (Early & 1.23E-05 8.63E+05 1.07E+01 Late)

Containment Bypass 8 5.94E-07 7.14E+05 4.24E-01 (ISLOCA, SGTR)

Total: 1.97E-05 1.186E+01

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

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.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Section 7.3.3.1 presents the Unit-1 population dose-rate data and calculations.

Section 7.3.3.2 presents the Unit-1 population dose-rate data and calculations.

7.3.3.1. Population Dose-Rate Calculations This section provides the Unit-1 population dose-rate data due to extending the ILRT interval including the OLB comparison with the CLB and PLB intervals, and the CLB and PLB interval.

Table 17 presents the Class 1 population dose-rate increase due to extending the ILRT interval.

Table 18 presents the Class 3a population dose-rate increase due to extending the ILRT interval.

Table 19 presents the Class 3b population dose-rate increase due to extending the ILRT interval.

Table 20 presents the Total (Class 1, 3a and 3b) population dose-rate increase due to extending the ILRT interval.

Equation 10 Percent Increase in Total Population Dose-Rate (PDR)

Equation 10 provides the percent increase in population dose-rate calculation.

Table 17 U1 - Class 1 PDR Increase Due to Extended Type A ILRT Intervals 3-Yrs 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (Baseline -Adjusted)

(person-rem) (person-rem)

(person-rem)

Class 1 PDR 1.68E-02 1.54E-02 1.44E-02 Class 1 PDR (OLBCLB/PLB) -1.39E-02 -2.39E-03 Class 1 PDR (CLBPLB) -9.96E-04 Table 18 U1 - Class 3a PDR Increase Due to Extended Type A ILRT Intervals 3-Yrs 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (Baseline -Adjusted)

(person-rem) (person-rem)

(person-rem)

Class 3a PDR 4.78E-03 1.59E-02 2.39E-02 Class 3a PDR (OLBCLB/PLB) 1.11E-02 1.91E-02 Class 3a PDR (CLBPLB) 7.98E-03 Table 19 U1 - Class 3b PDR Increase Due to Extended Type A ILRT Intervals 3-Yrs 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (Baseline -Adjusted)

(person-rem) (person-rem)

(person-rem)

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Class 3b PDR 1.19E-02 3.96E-02 5.94E-02 Class 3b PDR (OLBCLB/PLB) 2.77E-02 4.75E-02 Class 3b PDR (CLBPLB) 1.99E-02 Table 20 U1 -Total PDR Increase Due to Extended Type A ILRT Intervals 3-Yrs 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (Baseline -Adjusted)

(person-rem) (person-rem)

(person-rem)

PDRTotal(OLBCLB/PLB) 3.74E-02 6.43E-02 Class 1 + Class 3a + Class 3b PDRTotal(CLBPLB) 2.68E-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.

Equation 10 Percent Increase in Total Population Dose-Rate (PDR)

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

• 100%

Unit-1 Percent Increase in Total PDR (OLBCLB) = (3.74E-02 / 11.33)*100%

= 0.331%

Percentile Increase in Total (CLBPLB) = (2.68E-02 / 11.37)*100%

= 0.236%

Percentile Increase in Total PDR (OLBPLB) = (6.43E-02 / 11.33)*100%

= 0.568%

The following presents the Unit-2 population dose-rate data due to extending the ILRT interval including the OLB comparison with the CLB and PLB intervals, and the CLB and PLB interval.

Table 21 provides the Class 1 population dose-rate increase due to extending the ILRT interval.

Table 22 provides the Class 3a population dose-rate increase due to extending the ILRT interval.

Table 23 provides the Class 3b population dose-rate increase due to extending the ILRT interval.

Table 24 provides the Total (Class 1, 3a and 3b) population dose-rate increase due to extending the ILRT interval.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Table 21 U2 - Class 1 PDR Increase Due to Extended Type A ILRT Intervals Baseline -Adjusted 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (person-rem) (person-rem) (person-rem)

Class 1 PDR Increase 1.78E-02 1.63E-02 1.53E-02 Class 1 PDR (OLBCLB/PLB) -1.46E-03 -2.50E-03 Class 1 PDR (CLBPLB) -1.05E-03 Table 22 U2 - Class 3a PDR Increase Due to Extended Type A ILRT Intervals Baseline -Adjusted 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (person-rem) (person-rem) (person-rem)

Class 3a PDR Increase 5.01E-03 1.67E-02 2.51E-02 Class 3a PDR (OLBCLB/PLB) 1.17E-02 2.00E-02 Class 3a PDR (CLBPLB) 8.37E-03 Table 23 U2 - Class 3b PDR Increase Due to Extended Type A ILRT Intervals Baseline -Adjusted 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (person-rem) (person-rem) (person-rem)

Class 3b PDR Increase 1.25E-02 4.15E-02 6.24E-02 Class 3b PDR (OLBCLB/PLB) 2.91E-02 4.99E-02 Class 3b PDR (CLBPLB) 2.08E-02 Table 24 U2 Total PDR Increase Due to Extended Type A ILRT Intervals Baseline -Adjusted 10-Yrs (Current) 15-Yrs (Proposed)

ILRT Interval (person-rem) (person-rem) (person-rem)

PDRTotal(OLBCLB/PLB) 3.93E-02 6.74E-02 Class 1 + Class 3a + Class 3b PDRTotal(CLBPLB) 2.82E-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.

Unit-2 Percentile Increase in Total PDR (OLBCLB) = (3.93E-02 / 11.79)*100%

= 0.333%

Percentile Increase in Total PDR (CLBPLB) = (2.82E-02 / 11.83)*100%

= 0.238%

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Percentile Increase in Total PDR (OLBPLB) = (6.74E-02 / 11.79)*100%

= 0.572%

7.4 Step 4 - LERF and CCFP Changes In accordance with the methodology presented above, the LERF due to an ILRT 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 ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a larger release due to failure to detect a pre-existing leak during the extended window of vulnerability.

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 2.76E+04 person-rem) and the Class 3b is assumed to be 100 La (or 2.76E+05 person-rem). The method for defining the dose equivalent for allowable leakage (La) is developed in the EPRI report.[1 §4.3] This compares to a historical observed average of twice La. Therefore, the estimate is conservative.

Based on the EPRI guidance, only Class 3 sequences have the potential to result in large release is a pre-existing leak 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 Sequoyah and the change in LERF can be determined by the differences of the three test intervals. The EPRI guidance[2 §4.3] states that Class 3b sequences are considered the contributor to LERF associated with the Type A ILRT. Unit 1 a nd Unit 2, r espectively, summarizes the results of the LERF evaluation and the delta for the three 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 risk as resulting in increases below 1.0E-05/yr and 1.0E-06/yr, for CDF and LERF, respectively.[12 §2.4] Since the ILRT does not impact CDF, only LERF is relevant.

7.4.1 LERF Determination Equation 11 LERF Determination for Class 3b LERF = FREQClass_3b(CLB or PLB) - FREQClass_3b(OLB)

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Unit-1 LERF (OLBCLB) = FREQClass_3b (CLB) - FREQClass_3b(OLB)

= 1.43E-07/yr - 4.31E-08/yr

= 1.00E-07/yr LERF (CLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(CLB)

= 2.15E-07/yr - 1.43E-07/yr

= 7.19E-08/yr LERF (OLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(OLB)

= 2.15E-07/yr - 4.31E-08/yr

= 1.72E-07/yr Unit-2 LERF (OLBCLB) = FREQClass_3b (CLB) - FREQClass_3b(OLB)

= 1.51E-07/yr - 4.52E-08/yr

= 1.05E-07/yr LERF (CLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(CLB)

= 2.26E-07/yr - 1.51E-07/yr

= 7.55E-08/yr LERF (OLBPLB) = FREQClass_3b(PLB) - FREQClass_3b(OLB)

= 2.26E-07/yr - 4.52E-08/yr

= 1.81E-07/yr 7.4.2 Conditional Containment Failure Probability In accordance with the methodology[2, §5.1.4] presented above, the change in the Conditional Containment Failure Probability (CCFP) due to an ILRT interval extension is estimated as the difference in the CCFP values for the original and extended intervals.

Equation 12 Change in CCFP CCFP = [1 - (FREQClass_1(ILRT Interval) + FREQClass_3a(ILRT Interval) / CDFTotal] *100%

Unit-1 CCFP (OLB) = [1 - (FREQClass_1_U1_ADJ + FREQClass_3a_U1_OLB) / CDFTotal_U1] *100%

= [1 - (6.08E-06/yr + 1.73E-07/yr) / 2.07E-05/yr] *100%

= 69.776%

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

= [1 - (5.58E-06/yr + 5.76E-07/yr) / 2.07E-05/yr] *100%

= 70.261%

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

= [1 - (5.22E-06/yr + 8.65E-07/yr) / 2.07E-05/yr] *100%

= 70.608%

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Unit-2 CCFP (OLB) = [1 - (FREQClass_1_U2_ADJ + FREQClass_3a_U2_OLB) / CDFTotal_U2] *100%

= [1 - (6.44E-06/yr + 1.82E-07/yr) / 2.17E-05/yr] *100%

= 69.469%

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

= [1 - (5.92E-06/yr + 6.05E-07/yr) / 2.17E-05/yr] *100%

= 69.954%

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

= [1 - (5.54E-06/yr + 9.08E-07/yr) / 2.17E-05/yr] *100%

= 70.301%

Equation 13 %Change CCFP INCCCFP(For given ILRT Interval) = CCFP(ILRT Interval of Interest) - CCFPOLB Unit-1 CCFP Increase (OLBCLB) = CCFPCLB- CCFPOLB

= 70.261% - 69.776%

= 0.485%

CCFP Increase (OLBPLB) = CCFPPLB- CCFPOLB

= 70.608%% - 69.776%

= 0.832%

CCFP Increase (CLBPLB) = CCFPPLB- CCFPCLB

= 70.608%% - 70.261%

= 0.347%

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

= 69.954% - 69.469%

= 0.485%

CCFP Increase (OLBPLB) = CCFPPLB - CCFPOLB

= 70.301% - 69.469%

= 0.833%

CCFP Increase (CLBPLB) = CCFPPLB- CCFPCLB

= 70.301% - 69.954%

= 0.348%

7.4.3 Summary LERF - CCFP Table 25 and Table 26 summarizes the LERF and CCFP results for Unit-1 and Unit-2, respectively.

Table 25 Unit-1 Summary LERF - CCFP Risk Metric ILRT Interval Page E2-57 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval OLBCLB OLBPLB CLBPLB LERF 1.00E-07/yr 1.72E-07/yr 7.19E-08 CCFP 0.485% 0.832% 0.347%

Table 26 Unit-2 Summary LERF - CCFP ILRT Interval Risk Metric OLBCLB OLBPLB CLBPLB LERF 1.05E-07/yr 1.81E-07/yr 7.55E-08 CCFP 0.485% 0.833% 0.348%

8.0 Sensitivity Analyses The EPRI guidance for the analysis of extending the ILRT interval suggests using the liner corrosion sensitivity analysis performed by Calvert Cliffs.[14] 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, 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.[2, §5.1.5.1]

8.1 Liner Corrosion The analysis approach uses the Calvert Cliffs Nuclear Plant (CCNP) methodology[14] as modified by the EPRI guidance.[2] This methodology is an acceptable approach to incorporate the liner corrosion issue into the ILRT 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.

This analysis evaluates the sensitivity of risk impact results to assumptions in containment liner corrosion. The analysis approach uses the Calvert Cliffs Nuclear Plant (CCNP) methodology[14]

as modified by E PRI.[2] The methodology investigates how an age-related degradation mechanism can be factored into the risk impact associated with longer ILRT testing intervals.

The results of the analysis indicate that increasing the interval from three years (i.e., three in ten years) to fifteen years did not significantly increase plant risk of a large early release (Reference Table 30 and Table 31, for Unit 1 and Unit 2, respectively.

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.

The following approach is used to determine the change in likelihood, due to extending the ILRT interval, of detecting corrosion of the steel liner. This likelihood is used to determine the Page E2-58 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 for basemat concealed liner corrosion due to the lack of identified failures. (Table 27 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. 12 (Table 27 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 t he SQN Level 2 analysis has 1.0% leakage probability at 43 psia; whereas the design pressure is 26.7 psia.

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 27, 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. 13 (Table 27, 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 12 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.

13 During the 5.5 year data period used in the Calvert Cliffs analysis all liner corrosion events were detected through visual inspection. Sensitivity studies are included that evaluate total detection failure likelihoods of 5%

and 15%.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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).

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 27 Steps 2 and 3) 8.1.2 Differences in the Sequoyah Design from Calvert Cliffs 8.1.2.1 Structural Design The SQN containment design[24, §6.2.1.2] complies with NRC General Design Criteria 16.

The primary reactor containment is a f reestanding, 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.[24, §3.1]

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 [24, Table 6.2.1-1]

• Design Pressure, psig 12.0

• Design Temperature, 327°F

• Free Volume, 1,186,920 ft3

• Design & Maximum Allowable Leakage Rate, 0.25%/day

• Testing The SQN reactor containments meets NRC General Design Criteria (GDC) Criterion 52 and 53 with respect to integrated leak rate testing and inspections.[24, §3.1]

• Foundation & Bottom Liner Plate The primary containment foundation consists of a 9-foot thick circular reinforced concrete structural slab. [24, §3.8.5] The containment bottom liner plate is encased in concrete and is not visible. [24, §3.8.2]

• 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.[22]

The purpose of the general visual examination is to detect evidence of abnormal Page E2-60 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 ILRT, and during two subsequent outages preceding the next required ILRT.[22]

The containment liner has areas whereby 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.

8.1.3 Base Case Risk Assessment Table 27 summarizes the results obtained from the CCNP methodology using plant-specific data for Sequoyah.

Table 27 SQN Liner Corrosion Base-Case Risk Assessment Containment Cylinder and Containment Step Description Dome (85%) Basemat (15%)

Historical Liner Flaw Likelihood 14 Failure Data: Containment Location Specific Events: 2 Events: 0 1 Success Data: Based on 70 steel-lined containments and 5.5 years since the (Brunswick & North Anna) Assume a 1/2 Failure 10CFR50.55a requirements of periodic visual 2 / (70

• 5.5) = 5.19E-03 0.5 / (70

• 5.5) = 1.30E-03 inspection of cont. surfaces Year Failure Rate Year Failure Rate Aged Adjusted Liner Flaw Likelihood 1 2.05E-03 1 5.12E-04 During the fifteen-year interval, assume the failure 5-10 (ave) 5.19E-03 5-10 (ave) 1.30E-03 2 rate doubles every five years (14.9% increase per year). The average for the fifth to tenth year set to 15 1.43E-02 15 3.58E-03 the historical failure rate.

15 Year Ave = 6.44E-03 15 Year Ave = 1.61E-03 Increase in Flaw Likelihood Between Range  % Increase Range  % Increase Three and Fifteen Years 1 - 3 yrs 0.71 1 - 3 yrs 0.18 3

Uses aged adjusted liner flaw likelihood (Step 2), 1 - 10 yrs 4.06 1 - 10 yrs 1.03 assuming failure rate doubles every five years. 1 - 15 yrs 15 9.24 1 - 15 yrs 16 2.39 14 Containment location specific (consistent with the Calvert Cliffs analysis.[14]

15 The Calvert Cliffs analysis presents the delta between three and 15 years of 8.7% to utilize in the estimation of Page E2-61 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Containment Cylinder and Containment Step Description Dome (85%) Basemat (15%)

SQN DATA SQN DATA PSIA  % Failure PSIA  % Failure Likelihood of Breach in Containment 20 0.1 20 0.01 Given Liner Flaw 26.7 (ILRT) 0.4 26.7 (ILRT) 0.04 4 35 0.7 35 0.07

• Selective Approximation 17 43* 1.0 43* 0.10 50 18.8 50 1.88 60 44.2 60 4.42 83.7[13] 50.0 83.7 5.00 Assumed Value 18 1.0% 0.1%

Visual Inspection Detection Failure 5 10% 19 100% 20 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) 6 4.06% x 1.0% x 10% 1.03% x 0.1% x 100%

(Steps 3 x 4 x 5) 0.00410% 0.00103%

15-Year Test Interval (PLB) 15-Year Test Interval (PLB) 9.24% x 1.0% x 10% 2.39% x 0.1% x 100%

0.00924% 0.00239%

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.[2, §5.2.5.1]

16 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.[2, §5.2.5.1]

17 The 1% and 100% containment failure probabilities were selected by approximation on the containment capacity curve used in the Level 2 analysis.[16, Figure 5-1]

18 The failure probability of the cylinder and dome is assumed to be 1%, and the basement 0.1% as compared to 1.1% and 0.11% in the Calvert Cliffs analysis, this is conservative as SQNs calculated likelihood of a breach is less than 1% and 0.1% for the cylinder/dome and basemat, respectively. The failure probability at the ILRT pressure was determined by mathematical interpolation.

19 5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder but could be detected by ILRT). All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.

20 The containment basemat liner cannot be visually inspected.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 14 apply to both Sequoyah units.

Equation 14 Total Likelihood of Non-Detected Containment Leakage Total Likelihood of Non-Detected Containment Leakage (OLB) = 0.00071% + 0.00018% = 0.00089% or 8.90E-06 (CLB) = 0.00406% + 0.00103% = 0.00509% or 5.09E-05 (PLB) = 0.00925% + 0.00239% = 0.01164% or 1.16E-04 Equation 15 Liner Corrosion Non-LERF Frequency FREQNon-LERF = CDFBL - FREQClass 2 - FREQClass 3b(OLB) - FREQClass 8 Example (U1): = 1.88E-05/yr - 9.29E-08/yr - 4.31E-08/yr - 5.88E-07/yr

= 1.81E-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 28 and Table 29 for Unit 1 and 2, respectively.

• Per Table 11, the Unit-1 EPRI Class 3b frequency is 4.31E-08/yr.

• As shown in Table 5, the Sequoyah CDF associated with accidents that are not independently LERF or could never result in LERF (i.e., SERF, Class 6) is 2.07E-05/yr -

1.93E-06/yr = 1.88E-05/yr

• The OLB test interval data determined by Equation 14 (8.90E-06) is multiplied by the non-LERF probability determined by the example in Equation 15 (1.81E-05/yr), which results in an increase in LERF of 1.61E-10/yr (Table 28). 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 t est in 15 ye ars. A final adjustment can be made to address cases with containment spray operation; however, it is conservatively not addressed for Sequoyah and would not substantially alter the overall results.

Equation 16 Liner Corrosion - Increase in LERF Increase In LERF = FREQNon-LERF

• PNon-Detected leakage Table 28 Unit-1 Increase in LERF/yr Non-LERF Non- Increase In CDF/yr Class 3b/yr Frequency Case Class 2/yr Class 8/yr Detected LERF/yr Baseline OLB (Eq. 15) Leakage (Eq. 16)

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval (Eq. 14)

OLB 1.88E-05 9.29E-08 4.31E-08 5.88E-07 1.81E-05 8.90E-06 1.61E-10 CLB 1.88E-05 9.29E-08 1.43E-07 5.88E-07 1.81E-05 5.09E-05 9.19E-10 PLB 1.88E-05 9.29E-08 2.15E-07 5.88E-07 1.81E-05 1.16E-04 2.10E-09 Table 29 Unit-2 Increase in LERF/yr Non-LERF Non-Increase In CDF/yr Class 3b/yr Frequency Detected Case Class 2/yr Class 8/yr LERF/yr Baseline OLB Leakage (Eq. 15) (Eq. 16)

(Eq. 14)

OLB 1.97E-05 9.10E-08 4.52E-08 5.94E-07 1.90E-05 8.90E-06 1.69E-10 CLB 1.97E-05 9.10E-08 1.50E-07 5.94E-07 1.90E-05 5.09E-05 9.66E-10 PLB 1.97E-05 9.10E-08 2.26E-07 5.94E-07 1.90E-05 1.16E-04 2.21E-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 SQN Level 2 analysis has 1.0% leakage probability at 43 psia; whereas the design pressure is 26.7 psia. This methodology is consistent with the Calvert Cliffs methodology.[12]

In Equation 17 the increase in the CCFP (From Equation 12) due to assumed corrosion is calculated.

Equation 17 Increase in CCFP Due to Increase in Flaw Likelihood INCCCFP(ILRT Interval) = CCFP(ILRT Interval)

• 1.10% + CCFP(ILRT Interval)

Unit-1 INCCCFP_OLB = CCFPOLB

• 1.1% + CCFPOLB

= 0.698

• 1.1% + 0.698

= 0.705 INCCCFP_CLB = CCFPCLB

• 1.1% + CCFPCLB

= 0.703

• 1.1% + 0.703

= 0.710 INCCCFP_PLB = CCFPPLB

• 1.1% + CCFPPLB

= 0.706 *1.1% + 0.706

= 0.714 Unit 2 INCCCFP_OLB = CCFPOLB

• 1.1% + CCFPOLB

= 0.695

• 1.1% + 0.695

= 0.702 INCCCFP_CLB = CCFPCLB

• 1.1% + CCFPCLB

= 0.700

• 1.1% + 0.700 Page E2-64 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval

= 0.707 INCCCFP_PLB = CCFPPLB

• 1.1% + CCFPPLB

= 0.703

• 1.1% + 0.703

= 0.711 Equation 18 uses the CCFP calculated in Equation 12 and subtracts the Corrosion-Induced CCFP from Equation 17 to determine the increase in the CCFP due to corrosion.

Equation 18 CCFP Increase Due to Corrosion INCCCFP(ILRT Interval) = CCFP(With Corrosion) - CCFP(Without Corrosion)

Unit-1 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.705 - 0.698

= 0.00768 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.710 - 0.703

= 0.00773 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.714 - 0.706

= 0.00777 Unit 2 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.702 - 0.695

= 0.00764 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.707 - 0.700

= 0.00769 INCCCFP_OLB = CCFPOLB(With) - CCFPOLB(Without)

= 0.711 - 0.703

= 0.00773 8.1.6 Summary of Base Case and Corrosion Sensitivity Cases This contribution of corrosion-induced LERF likelihood determined in section 8.1.4 is added to the Class 3b LERF cases and the sensitivity analysis performed. Table 30 and Table 31 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 ILRT interval:

• Base Case, which represent 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.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval 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 in percentage points.

• Class 3b LERF is provided and indicated the accident class 3b frequency as well as the change in the class 3b frequency below.

• The row title 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 ILRT 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 ILRT frequency from the CLB. The difference between the non-corrosion and corrosion-cases is provided.

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Table 30 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 Without Without With Corrosion Without Corrosion With Corrosion With Corrosion Class Corrosion Corrosion Per- Person- Person- Per- Person- Person- Per-Person- Per-Rem Rem Rem Rem Rem Rem Rem Frequency Rem Per Frequency Rem Per Frequency Frequency Frequency Frequency Per Per Per Per Per Per Per Year Year Year Year Year Year Year Year Year 1 6.08E-06 1.67E-02 6.08E-06 1.68E-02 4.43E-07 5.58E-06 1.54E-02 5.58E-06 1.51E-02 -2.54E-06 5.22E-06 1.44E-02 5.22E-06 1.38E-02 -5.80E-06 2 9.29E-08 6.96E-01 9.29E-08 6.96E-01 N/A 9.29E-08 6.96E-01 9.29E-08 6.96E-01 N/A 9.29E-08 6.96E-01 9.29E-08 6.96E-01 N/A 3a 1.73E-07 4.78E-03 1.73E-07 4.78E-03 N/A 5.76E-07 1.59E-02 5.76E-07 1.59E-02 N/A 8.65E-07 2.39E-02 8.65E-06 2.39E-02 N/A 3b 4.31E-08 1.19E-02 4.32E-08 1.19E-02 4.43E-05 1.43E-07 3.96E-02 1.44E-07 3.98E-02 2.54E-04 2.15E-07 5.94E-02 2.17E-07 6.00E-02 5.80E-04 7 1.18E-05 1.02E+01 1.18E-05 1.02E+01 N/A 1.18E-05 1.02E+01 1.18E-05 1.02E+01 N/A 1.18E-05 1.02E+01 1.18E-05 1.02E+01 N/A 8 5.88E-07 4.20E-01 5.88E-07 4.20E-01 N/A 5.88E-07 4.20E-01 5.88E-07 4.20E-01 N/A 5.88E-07 4.20E-01 5.88E-07 4.20E-01 N/A CDF/

Total 1.88E-05 1.13E+01 1.88E-05 1.13E+01 4.47E-05 1.88E-05 1.14E+01 1.88E-05 1.14E+01 2.51E-04 1.88E-05 1.14E+01 1.88E-05 1.14E+01 5.74E-04 Dose Dose Dose Dose Dose Dose (per-rem/yr) 3.75E-02 (per-rem/yr) 3.74E-02 6.43E-02 6.42E-02 N/A N/A (per-rem/yr) (per-rem/yr)

Rate

%Increase 0.33% %Increase 0.33% %Increase 0.57% %Increase 0.57%

CCFP 69.78% 70.54% 70.26% 71.03% 70.61% 71.38%

CCFP N/A N/A 0.485% 0.490% 0.832% 0.841%

Class 4.31E-08 4.32E-08 1.43E-07 1.44E-07 2.15E-07 2.17E-07 3b LERF Class 3b & Non-1.61E-10 N/A 9.19E-10 N/A 2.10E-09 Corrosion LERF LERF (from base case of 3 per 10 years) 1.00E-07 1.01E-07 1.72E-07 1.74E-07 LERF from 1 per 10 years N/A 7.19E-08 7.31E-08 Page E2-67 of 77

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Sequoyah - Units 1 & 2 PRA Evaluation for Permanent Extension to the Containment Type A ILRT Interval Table 31 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 Without With Corrosion Without Corrosion With Corrosion With Corrosion Class Corrosion Corrosion Per- Person- Person- Per- Person- Person- Per-Person- Per-Rem Rem Rem Rem Rem Rem Rem Frequency Rem Per Frequency Rem Per Frequency Frequency Frequency Frequency Per Per Per Per Per Per Per Year Year Year Year Year Year Year Year Year 1 6.44E-06 1.78E-02 6.44E-06 1.78E-02 -4.66E-07 5.92E-06 1.63E-02 5.91E-06 1.61E-02 -2.66E-06 5.54E-06 1.53E-02 5.53E-06 1.47E-02 -6.09E-06 2 9.10E-08 6.82E-01 9.10E-08 6.82E-01 N/A 9.10E-08 6.82E-01 9.10E-08 6.82E-01 N/A 9.10E-08 6.82E-01 9.10E-08 6.82E-01 N/A 3a 1.82E-07 5.01E-03 1.82E-07 5.01E-03 N/A 6.05E-07 1.67E-02 6.05E-07 1.67E-02 N/A 9.08E-07 2.51E-02 9.08E-07 2.51E-02 N/A 3b 4.52E-08 1.25E-02 4.54E-08 1.25E-02 4.66E-05 1.50E-07 4.15E-02 1.51E-07 4.18E-02 2.66E-04 2.26E-07 6.24E-02 2.28E-07 6.30E-02 6.09E-04 7 1.23E-05 1.07E+01 1.23E-05 1.07E+01 N/A 1.23E-05 1.07E+01 1.23E-05 1.07E+01 N/A 1.23E-05 1.07E+01 1.23E-05 1.07E+01 N/A 8 5.94E-07 4.24E-01 5.94E-07 4.24E-01 N/A 5.94E-07 4.24E-01 5.94E-07 4.24E-01 N/A 5.94E-07 4.24E-01 5.94E-07 4.24E-01 N/A CDF/

Total 1.97E-05 1.18E+01 1.97E-05 1.18E+01 4.61E-05 1.97E-05 1.18E+01 1.97E-05 1.18E+01 2.64E-04 1.97E-05 1.19E+01 1.97E-05 1.19E+01 6.03E-04 Dose Dose Dose Dose Dose Dose (per-rem/yr) 3.93E-02 (per-rem/yr) 3.92E-02 6.74E-02 6.74E-02 N/A N/A (per-rem/yr) (per-rem/yr)

Rate

%Increase 0.33% %Increase 0.33% %Increase 0.57% %Increase 0.57%

CCFP 69.47% 70.23% 69.95% 70.72% 70.30% 71.07%

CCFP N/A N/A 0.485% 0.490% 0.833% 0.842%

Class 4.52E-08 4.54E-08 1.50E-07 1.51E-07 2.26E-07 2.28E-07 3b LERF Class 3b & Non-1.69E-10 N/A 9.66E-10 N/A 2.21E-09 Corrosion LERF LERF (from base case of 3 per 10 years) 1.05E-07 1.06E-07 1.81E-07 1.83E-07 LERF from 1 per 10 years N/A 7.55E-08 7.67E-08 Page E2-68 of 77

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PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK 8.1.7 Liner Corrosion Sensitivity Conclusion As shown in Table 30 and Table 31 for Unit 1 a nd Unit 2, r espectively, the inclusion of corrosion does not result in an increase in LERF sufficient to invalidate the baseline analysis and the overall impact is negligible.

9.0 Evaluation of External Events In this step, the potential contribution from external events is estimated as a result of increasing the ILRT interval. Due to lack of detailed Level 2 P RA modeling availability for external events, their potential contribution is limited to a conservative estimate of the change in LERF associated with the ILRT interval extension. Seismic and 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 seismic and internal fire events bound the risk contribution from other external events. Both seismic and internal fires were examined but the analyses contained conservative assumptions.

External events were evaluated in the Sequoyah Individual Plan Examination of External Events (IPEEE).[13] The IPEEE was a one-time review of external hazards risk and was limited in its purpose to the identification of potential plant vulnerabilities and an understanding of severe accident risk.

9.1 Seismic Analysis SQN performed a S eismic Margins Assessment (SMA) following the IPEEE guidance of NUREG-1407 and the seismic margins methodology of E PRI NP-6041-SL.[20, §1.3] The SMA approach is a d eterministic and conservative evaluation that does not calculate risk on a probabilistic basis. Therefore, its results should not be compared directly with the best-estimate internal events results.

The conclusions of the SQN IPEEE seismic margin analysis[20] and subsequent analyses/evaluations are as follows:

• The equipment reviewed for SQN during the systematic evaluation of the seismic event proved to be overall rugged in nature and of a sufficient capacity to provide assurance of continued functionality for the Review Level Earthquake (RLE).

• This systematic evaluation of the seismic event performed by the Seismic Margins Method has provided adequate evidence of the ability of SQN to resist a significant seismic event up to the RLE and be able to initiate a safe shutdown of the unit.

Therefore, the potential for core damage, containment failures, or off site releases are considered acceptably low. No specific vulnerabilities to the seismic event were noted other than a few specific areas of improvement. The plant improvements are identified in NUREG-1742 as replacement of MCC anchorages; upgrade of RHR heat exchanger anchorages; and corrective change to eliminate interactions. All of these improvements have been implemented.

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SUMMARY

NOTEBOOK

• The original evaluation determined that the RHR heat exchangers had a High Confidence, Low Probability of Failure (HCLPF) of 0.27g. M odifications were made and the heat exchangers are no longer considered low capacity components. No other unique decay heat removal vulnerabilities to seismic events were found.[25, E.1.3.4]

• As originally evaluated, assuming a ground level RLE of 0.3g, t he overall plant HCLPF capacity at SQN was determined to be at least 0.27g. I n response to an NRC request for additional information (RAI), certain components were reevaluated assuming a RLE defined by a NUREG/CR-0098 spectral shape anchored to 0.30g a t rock. The limiting recomputed component HCLPF values range from 0.23g to 0.29g .[25, E.1.3.4]

• As originally evaluated, in the Sequoyah License Extension work a t echnical review was conducted of the HCLPF for those components that during the IPEEE were less than 0.3g.

Each of the 12 noted components noted in the IPEEE were re-analyzed. One component, the 480v Shutdown Transformer required a minor modification to the anchorage which is complete. Therefore, for this analysis it is assumed that Sequoyah can mitigate the consequence of an earthquake up to 0.3g intensity.

In section 9.2.1, the potential contribution from seismic events is estimated. Due to lack of detailed Level 2 PRA modeling availability for seismic events, their potential contribution is limited to a conservative estimate of the change in LERF associated with the extension of the ILRT interval.

The Seismic CDF for Sequoyah is obtained from GI-199.[31, Table D-1] This table uses the 2008 Seismic Hazard Curve data determined by t he United States Geological Survey (USGS). The Simple Average was selected for the representative value for Sequoyah which corresponds to a seismic CDF of 2.90E-05/yr.

9.2 Internal Fires Analysis The findings contained in NUREG-1742[26] 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 t his 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 Sequoyah internal fire analysis was performed in 1995 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. Most of the SQN fire areas were screened in the early screening phase. The CDF of the areas in the final phase of screening totaled 1.56E-05/yr.[25, §E.1.3.2]

A revised fire IPEEE was developed in response to NRC RAIs. The subsequent evaluation was able to screen more areas because of additional walkdowns and cable routing information which resulted in more credit for feed and bleed cooling and the use of fire severity factors. The conclusion of the analysis was that all rooms were screened from further consideration and it was Page E2-70 of 77

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PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK confirmed that there are non fire-induced vulnerabilities associated with the continued operation of Sequoyah. The total CDF of the areas remaining in the final phase of screening for the this revision was 5.83E-06/yr.[25, §E.1.3.2]

Conservatively, the 1.56E-05/yr frequency is used in this analysis. However, in recent years the Sequoyah 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).

9.3 External Events Contribution External events contribution to CDF is determined by summing the seismic-induced CDF with the fire-induced CDF. The contribution from other external hazards such as high winds, transportation, etc., are relatively insignificant to the consequences of seismic and internal fire events. Sequoyah performed the screening described in Supplement 4 t o Generic Letter 88-20 and NUREG-1407 to address other external hazards. Because SQN was designed prior to the 1975 Standard Review Plan (SRP) the approach taken was to review the design basis and compare them to the SRP requirements. Any changes to the plant since the design analyses were performed were also reviewed to verify compliance with SRP criteria. It was found that no vulnerabilities exist for other external events which are not within the screening threshold of the SRP. The IPEEE evaluation revealed that the plant meets the 1975 SRP criteria for these external events and no recommendations for plant improvements.[25, § E.1.3.3] In recent years the Sequoyah plant has reevaluated hydrology issues, identified vulnerabilities and performed modifications to eliminate those vulnerabilities.

9.3.1 External Events Contribution to CDF The following calculation and results are applicable for both Sequoyah units.

Equation 19 External Events Contribution to CDF CDFEE = CDFSeismic + CDFFire

= 2.90E-05/yr + 1.56E-05/yr

= 4.46E-05/yr 9.3.2 External Events Contribution to LERF Equation 20 the CDF due to external events is multiplied by the probability of a Class 3b event.

The product is the frequency for Class 3b. This calculation applies to both Unit 1 and Unit 2.

Equation 20 External Events Impact on the Class 3b Frequency FREQClass 3b(OLB-EE) = CDFEE

• PClass 3b

= 4.46E-05/yr

• 0.002294

= 1.02E-07/yr Page E2-71 of 77

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SUMMARY

NOTEBOOK 9.3.3 External Events Contribution to LERF for ILRT Interval Extension This section characterizes the change in risk associated with external events. The following tables are provided in this section.

Equation 21 presents the external events impact on the Class 3b frequency for the extended ILRT test intervals Equation 22 presents the external events impact on the change in LERF (3b only) from extended ILRT test intervals Table 32 presents the results for external events contribution to the ILRT interval extensions Equation 23 presents the total external events impact on LERF from extending ILRT test interval FREQLERF(EE)_Total_OLB_CLB or PLB = FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE_CLB-EE or PLB-EE) where; U1 FREQClass_2, Class_7(LERF), Class_8 = 2.98E-06/yr (Table 5)

U2 FREQClass_2, Class_7(LERF), Class_8 = 3.05E-06/yr (Table 5)

FREQClass_3b(OLB-EE) = 1.02E-07/yr (Table 32)

FREQClass_3b(CLB-EE) = 3.41E-07/yr (Table 32)

FREQClass_3b(PLB-EE) = 5.11E-07/yr (Table 32)

Unit 1 FREQLERF(EE)_Total_OLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE)

= 2.98E-06/yr + 1.02E-07/yr

= 3.08E-06/yr FREQLERF(EE)_Total_CLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(CLB-EE)

= 2.98E-06/yr + 3.41E-07/yr

= 3.32E-06/yr FREQLERF(EE)_Total_PLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(PLB-EE)

= 2.98E-06/yr + 5.11E-07

= 3.49E-06/yr Unit 2 FREQLERF(EE)_Total_OLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE)

= 3.05E-06/yr + 1.02E-07/yr

= 3.15E-06/yr FREQLERF(EE)_Total_CLB_U2 = U2 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(CLB-EE)

= 3.05E-06/yr + 3.41E-07/yr

= 3.39E-06/yr FREQLERF(EE)_Total_PLB_U2 = U2 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(PLB-EE)

= 3.05E-06/yr + 5.11E-07

= 3.56E-06/yr Table 33 presents the Unit 1 upper bound on the external events LERF to the ILRT interval extensions Table 34 presents the Unit 2 upper bound on the external events LERF to the ILRT interval extensions In Equation 21 the Class 3b frequency determined by multiplying the by the ILRT interval factor, 3.33 for 1-in-10 years, and 5.0 for 1-in-15 years. The results apply to both units.

Equation 21 EE Impact on the Class 3b Frequencies for Extended ILRT Intervals FREQEE-Class 3b(xx-yr) =FREQClass 3b(OLB-EE)* MULTCLB (or MULTPLB)

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SUMMARY

NOTEBOOK FREQClass 3b(CLB-EE) = 1.02E-07/yr

• 3.33

= 3.41E-07/yr FREQClass 3b(PLB-EE) = 1.02E-07/yr

• 5.0

= 5.11E-07/yr Given the values calculated by Equation 21, the LERF increases for the extended ILRT intervals due to the external events contribution is determined by Equation 22. These values apply to both units.

Equation 22 External Events Contribution to a Change in LERF LERFEE(OLBCLB) = FREQClass 3b(CLB-EE) - FREQClass 3b(OLB-EE)

= 3.41E-07/yr - 1.02E-07/yr

= 2.38E-07/yr LERFEE(OLBPLB) = FREQClass 3b(PLB-EE) - FREQClass 3b(OLB-EE)

= 5.11E-07/yr - 1.02E-07/yr

= 4.09E-07/yr Table 32 External Events Contribution to Risk for ILRT Interval Extension FREQEE-LERF/yr LERF Increase/yr CDFEE/yr PClass 3b OLB-EE CLB-EE PLB-EE OLB-EE CLB-EE OLB-EE PLB-EE 4.46E-05 0.00229 1.02E-07 3.41E-07 5.11E-07 2.38E-07 4.09E-07 Only the Class 3b events are affected by an increase in the ILRT 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, the application will be considered if it can be reasonably shown that the total LERF is less than 1.0E-05/rx-yr, also known as region II on Figure 5 of R.G. 1.174.[12, §2.4]

Equation 23 Total External Events Impact on LERF from Extending ILRT Interval FREQLERF(EE)_Total_OLB_CLB or PLB = FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE_CLB-EE or PLB-EE) where; U1 FREQClass_2, Class_7(LERF), Class_8 21 = 2.98E-06/yr (Table 5)

U2 FREQClass_2, Class_7(LERF), Class_8 = 3.05E-06/yr (Table 5)

FREQClass_3b(OLB-EE) = 1.02E-07/yr (Table 32)

FREQClass_3b(CLB-EE) = 3.41E-07/yr (Table 32) 21 As a comparison, conservatively assuming that LERF for external events is equal to that of internal events (Class 2, 7(LERF Only), and 8) gives a Total LERF of 5.96E-06/yr and 6.10E-06/yr, Unit 1 and Unit 2, respectively.

Comparing the two methods above both meet the R.G. 1.174 criteria for allowing a small change in delta LERF with respect to total LERF.[12]

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NOTEBOOK FREQClass_3b(PLB-EE) = 5.11E-07/yr (Table 32)

Unit 1 FREQLERF(EE)_Total_OLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE)

= 2.98E-06/yr + 1.02E-07/yr

= 3.08E-06/yr FREQLERF(EE)_Total_CLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(CLB-EE)

= 2.98E-06/yr + 3.41E-07/yr

= 3.32E-06/yr FREQLERF(EE)_Total_PLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(PLB-EE)

= 2.98E-06/yr + 5.11E-07

= 3.49E-06/yr Unit 2 FREQLERF(EE)_Total_OLB_U1 = U1 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(OLB-EE)

= 3.05E-06/yr + 1.02E-07/yr

= 3.15E-06/yr FREQLERF(EE)_Total_CLB_U2 = U2 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(CLB-EE)

= 3.05E-06/yr + 3.41E-07/yr

= 3.39E-06/yr FREQLERF(EE)_Total_PLB_U2 = U2 FREQClass_2, Class_7(LERF), Class_8 + FREQClass_3b(PLB-EE)

= 3.05E-06/yr + 5.11E-07

= 3.56E-06/yr Table 33 U1 - Upper Bound on All LERF EPRI Accident Classes FREQ/yr LERF Increase/yr Hazard OLB CLB PLB OLB CLB OLB PLB 22 External 3.08E-06 3.32E-06 3.49E-06 2.38E-07 4.09E-07 Events Internal 3.02E-06 3.12E-06 3.19E-06 1.00E-07 1.72E-07 Events Combined 6.10E-06 6.44E-06 6.68E-06 3.39E-07 5.81E-07 Table 34 U2 - Upper Bound on All LERF Contributors EPRI Accident Classes FREQ/yr LERF Increase/yr Hazard OLB CLB PLB OLB CLB OLB PLB External 3.15E-06 3.39E-06 3.56E-06 2.38E-07 4.09E-07 Events Internal 3.09E-06 3.20E-06 3.27E-06 1.05E-07 1.81E-07 Events Combined 6.25E-06 6.59E-06 6.84E-06 3.44E-07 5.90E-07 22 External Events LERF is determined by adding the contribution from Class 3b for the three test intervals, and adding the values to the Class 2 and Class 8 values. Only Class 3b is affected by the ILRT interval, therefore, the values for Class 2 and 8 remain static across the three test intervals.

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NOTEBOOK 10.0 Results/Conclusion NEI 94-01, Revision 3-A,[5] 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[11] and includes provisions for permanently extending Type A intervals to 15 years. Based on t he 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 ILRT Type A extension to one 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 36.

Table 35 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 36 presents the results of the calculations and the applicability to this application Table 35 Acceptance Criteria Figure of Merit Acceptance Criteria (Increase Above Baseline) Source LERF <1.0E-05/rx-yr RG 1.174 §2.4[12]

LERF <1.0E-06/rx-yr RG 1.174 §2.4[12]

CCFP 23 1.5% NEI 94-01 R2 §2.2[4]

Dose increase 1.0 person-rem/yr or 1% of the total Dose (person-rem) EPRI 1018243 APP H[2]

baseline dose, whichever is less restrictive In the discussions that follow the maximum results for Unit 1 and Unit 2 is provided for LERF.

CCFP and Dose. Unit 2 bounds Unit 1 in all three figures of merit, therefore only that value is presented in the discussions. However, Table 36 provides the data for both units.

10.1 Results Discussion - LERF Regulatory Guide 1.174[12] 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 Damage Frequency (CDF); therefore, there is no c hange 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 Sequoyah, the analysis included the estimated contribution from external events in addition to the internal events analysis. Table 33 and Table 34 summarizes the maximum LERF for Sequoyah which is estimated to be 5.90E-07/yr, AND the maximum upper bound total LERF (Including External Events) 6.84E-06/yr. Both results are within the acceptable bands for a small change in risk according to R.G. 1.174. Table 36 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 23 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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SUMMARY

NOTEBOOK change from the original licensing basis (OLB) of 3 test 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 1009325[1] a maximum conditional containment failure probability (CCFP) increase for Sequoyah from the OLB to the PLB is 0.842% which includes the increased contribution due to aging and corrosion affects. Revision 2-A of the EPRI Report characterizes an increase in the CCFP of 1.5% as very small.[2, §2.2] This is consistent with the NRC Final Safety Evaluation for NEI 94-01[4] and EPRI Report 1009235.

Therefore, this increase is judged to be small. Table 36 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 ILRT interval to 1-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 6.74E-02 person-rem/yr which corresponds to increase of 0.57% is calculated for Sequoyah. This value is based on internal events only. EPRI Report 1009325, Revision 2-A[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 ILRT interval extension to 15 years. Therefore, Sequoyah meets both metrics. This is consistent with the NRC Final Safety Evaluation for NEI 94-01[4] and EPRI Report 1009235.

Table 36 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 36 include the increased risk from corrosion.

Table 36 Results Table and Applicability Determination Unit 1 Acceptable for Metric Value Acceptance Criteria Application?

LERFIE-Total 2.98E-06/yr

<1.0E-05/rx-yr Yes LERFTotal(IE & EE) 6.68E-06/yr LERFTotal(OLBCLB) 3.39E-07/yr

<1.0E-06/rx-yr Yes LERFTotal(OLBPLB) 5.81E-07/yr CCFP(OLBCLB), Inc. Corrosion 0.490%

1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.841%

DOSE(OLBCLB) 3.75E-02 per-rem/yr

<1.0 person-rem/yr or <1% of total DOSE(OLBPLB) 6.43E-02 per-rem/yr Yes dose, whichever is less restrictive.

%DOSE(OLBCLB) 0.33%

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NOTEBOOK

%DOSE(OLBPLB) 0.57%

Unit 2 LERFIE-Total 3.05E-06 <1.0E-05/rx-yr Yes LERFTotal(IE & EE) 6.84E-06 LERFTotal(OLBCLB) 3.44E-07 <1.0E-06/rx-yr Yes LERFTotal(OLBPLB) 5.90E-07 CCFP(OLBCLB), Inc. Corrosion 0.490% 1.5% Yes CCFP(OLBPLB), Inc. Corrosion 0.842%

DOSE(OLBCLB) 3.93E-02 per-rem/yr DOSE(OLBPLB) 6.74E-02 per-rem/yr <1.0 person-rem/yr or <1% of total Yes

%DOSE(OLBCLB) 0.33% dose, whichever is less restrictive.

%DOSE(OLBPLB) 0.57%

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ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 RESOLUTIONS TO PROBABILISTIC RISK ASSESSMENT PEER REVIEW TEAM FACTS AND OBSERVATIONS

Subject:

Application to Revise Technical Specification 6.8.4.h, "Containment Leakage Rate Testing Program," (SQN-TS-14-03)

Page E3-1 of 87 Resolutions to F&Os[10]

Finding Level F&Os F&O Number F&O Details 1-4 MDN-000-000-2010-0203 does not document an assessment of the impact of flooding events on existing HFEs carried over from the internal events scenario used to represent the flooding event.

(This F&O originated from SR IFQU-A6)

Associated SR(s)

IFQU-A6 Basis for Significance Although this is a documentation issue, it is important to an understanding of the results and to show that the technical element is satisfied.

Possible Resolution Document a process for assessment of the impact of the flood scenarios on existing HFEs from the internal events PRA sequences used to represent the flood scenarios.

EPRI 1019194 Section 7.3 describes the types of HFE adjustments to be considered.

Response

To address human actions and their modification due to flooding events Section 9.3 was added to the document.

Section 9.3 addresses the changes to the human actions in the model by accounting for:

1. Human actions that are influenced by HRA actions, these are events that occur within an hour of flood initiation.

2. Human actions that are failed due flooding.

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F&O Number F&O Details 1-7 Dependency analysis was performed for the post-initiator HEPs using the EPRI HRA Calculator. However, several issues were identified including:

1) Use of the same cue for two actions can result in conservative dependency values. For example, the use of the same cue for actions HARR1 and AFWOP3 resulted in complete dependency between the actions. However, review of the cues indicated that the cue for AFWOP3 should be different than that for HARR1.

2) Inconsistent entry of the timing information creates results that may appear invalid. For example, the timing entries for actions HARR2 and AFWOP3 make it appear that core damage as a result of failure of HARR2 would occur before the cue for AFWOP3 is received. Discussion revealed that the Tsw for HARR2 is based on the time at which the RWST would empty rather than core damage as stated in the HRA Calculator.

3) Inclusion of screening HFEs in the dependency analysis can result in errors. The screening HEPs do not have information that is necessary for the dependency analysis (e.g., timing inputs). This can result in the wrong event being treated as the independent event in the combination. For example, review of dependency combination 41 shows that the dependency analysis treats HACD1 as the first or independent HFE in the combination and AFWOP5 as following HACD1. This results in a joint HEP of 1.0 based on complete dependency. However, the description of HFE HACD1, Perform cooldown with main feedwater following AFW failure, indicates that AFWOP5 should be the first event. This would result in a joint HEP of 2.9E-03.

4) The dependency level of the cognitive recoveries were not entered in the HRA Calculator database for the post-initiators. This requires manual entry by the analyst and does not default to the recommended dependence level. Failure to enter this information may underestimate or overestimate the HEP depending on the applicable dependence level.

Some of these items were corrected during the review but they are documented in an F&O due to the need to evaluate the extent of the condition.

(This F&O originated from SR HR-G7)

Associated SR(s)

HR-G7 Basis for Significance Incorrect assignment of cues, timing, and resource entries can result in incorrect dependency analysis results.

Possible Resolution

1) Review the cues, timing assumptions, and resource requirements for significant HEPs to ensure that the factors are correctly assessed in the dependency analysis.

2) For combinations where the timing indicates Tsw is reached for the first action before the cue for the second is received, document the basis for Page E3-3 of 87

acceptability of the dependent combination.

3) Ensure all information affecting the dependency analysis is entered into the HRA Calculator for the screening HFEs to ensure they are treated correctly in the dependency analysis.

4) Ensure that the dependence level between cognitive actions and applicable recoveries is set in the HRA Calculator database.

Response

1. Cue for AFWOP3 has been updated to correct cue. Review has been performed for all remaining actions to determine if any additional cues need to be updated. This review verified the accuracy of HRA cues and updated six of the identified cues.

2. The end point for Tsw is an irreversible damage state. For HARR2, this irreversible damage state is the loss of all ECCS pumps when the RWST is depleted and autoswap has failed. This is the correct irreversible damage state as the operator does not have until core damage to perform that action if the pumps fail when their suction source runs dry. The dependency analysis was reviewed for overlapping timeframes.

3. Screening value HEPs were removed from the database if there values were set to 1.0. The HEPs that were originally in the model were no longer required and were deleted from the fault tree.

4. This has been corrected for all of the actions in the SQN HRA.

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F&O Number F&O Details 1-8 MDN-000-000-2010-0203 Section 9.5 only addresses quantification and results for CDF. There is no discussion of LERF for the flooding scenarios or documentation indicating that the flood scenarios were reviewed to determine if they would have an impact on the Level 2 CETs. The linked fault tree model should have the capability to produce LERF results, but this had not been done at the time of the review. In addition, there was no discussion in the Level 2 Notebook (MDN-000-000-2010-0206) that indicates the results include the internal flood scenarios.

(This F&O originated from SR IFQU-A10)

Associated SR(s)

IFQU-A10 Basis for Significance No LERF results for internal flooding scenarios was provided for review.

Possible Resolution

1) Document a review of the top events in the Level 2 model to confirm that there are no unique flooding impacts that affect the CETs.

2) Document the LERF results for the internal flood scenarios similar to the results for other initiators in Section 11 of MDN-000-000-2010-0206. This can be done in the flood notebook or the Level 2 notebook, but, if done in the Level 2 notebook this should be referenced in the Internal Flooding Analysis notebook.

Response

The internal flooding calculation was revised to add Section 10 (Results Analysis for Large Early Release Frequency).

Section 10.1 addresses the eighteen questions concerning LERF and their impact.

Section 10.3 and 10.4 address the LERF results due to flooding To address the additional information the following Appendices were added to the model:

Appendix Q - Significant Cutset Review for Large Early Release Appendix R - Non-Significant Cutset Review for Large Early Release Appendix S - Importance Reports for Large Early Release Page E3-5 of 87

F&O Number F&O Details 1-10 MDN-000-000-2010-0206 Section 5.6 notes that credit was taken for scrubbing of releases from a ruptured SG. However, the technical justification for this credit needs to be strengthened. The current basis compares the zero power collapsed level to the top of the SG tubes. However, ES-3.1, Post-SGTR Cooldown Using Backfill allows the level in the ruptured SG to be between 20% narrow range and 75%

narrow range during the cooldown (Step 7). The expected levels during SGTR recovery should be used to justify the scrubbing credit.

It also appears that the analysis implicitly assumes that if FW will be applied to the ruptured SG if FW is available. No consideration of operator failure to provide FW flow to the ruptured generator is included in the analysis.

(This F&O originated from SR LE-C4)

Associated SR(s)

LE-C4 LE-C13 LE-E3 Basis for Significance The technical basis of the credit for scrubbing of SGTR releases does not consider the levels allowed in the EOPs.

Possible Resolution Revise the justification in MDN-000-000-2010-0206 Section 5.6 to include consideration of the SG levels maintained during recovery using the applicable EOPs.

Response

The documentation has been updated to include a discussion of the water levels above the steam generator tubes during tube rupture recovery actions. These water levels (between 4.7 and 9.8 feet) should be sufficient to take credit for fission product scrubbing. This analysis assumes that the operator is successful in providing feedwater flow to the ruptured steam generator.

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F&O Number F&O Details 1-11 The total LERF is compared with other Westinghouse 4-loop plants and with other Ice Condenser plants. However, there is no comparison at the level of significant contributors or plant damage states. Without the contributor information, it is not really possible to determine how similar the LERF results are to other plants.

(This F&O originated from SR LE-F2)

Associated SR(s)

LE-F2 Basis for Significance There is no review of the contributors to LERF with the results for similar plants to ensure that plant-specific modeling choices have not skewed the results.

Possible Resolution Document a comparison of the LERF results to plants of similar design at the significant contributor and PDS levels (similar to Tables 11-1 and 11-3).

Response

The documentation has been updated to include comparisons by initiating event for several other PWRs in Table 11-7.

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F&O Number F&O Details 1-14 Demand data is obtained directly from the plant process computer for most components, as described in Section 7.3 of the data notebook (MDN-000-000-2010-0202). The status change information from the computer is filtered and used to determine the number of demands.

The use of automatic data collection, however, means that start and run events that occur in all modes of operation are included. In addition, post-maintenance test starts are also included in the data set. This is identified as a source of uncertainty in the sensitivities and uncertainties notebook (MDN-000-000-2010-0209) and a specific set of sensitivity studies were performed that assumed that various numbers of successful starts were invalid. The results show that the impact on CDF is relatively small, unless the number of successful starts is overestimated by a large amount. However, this SR is explicit in its requirement to not count post-maintenance test events.

(This F&O originated from SR DA-C6)

Associated SR(s)

DA-C6 Basis for Significance This is considered to be a finding since a specific technical requirement of the SR is not met.

Possible Resolution To comply with this SR, the post-maintenance test starts (following a component failure) should be removed. It would also be more correct to also screen out component demands that occur during shutdown periods (e.g., by filtering out data based on the date of the event).

Response

The work orders for the components that were credited for success in the data analysis were reviewed to discover the number of post maintenance tests that were performed on the components. Table 15 was added to document the number of post maintenance tests that were removed from the analysis.

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F&O Number F&O Details 1-15 The super initiator "general transient" may overlook certain differences among its contributors. For example, the impact of specific IEs like LOSP and Loss of DC that may prevent PORV operation and challenge the Pressurizer Safeties do not appear to be captured.

In addition, failure to provide a separate event tree for SBO may overestimate the success of power recovery by not addressing the operation of systems such as charging and AFW following power recovery.

(This F&O originated from SR AS-A10)

Associated SR(s)

AS-A10 AS-B1 SC-B3 Basis for Significance The accident sequences do not contain sufficient detail to capture important system requirements and required operator interactions for all initiating events.

Possible Resolution

1) Subdivide the General Transients event tree to better represent the unique challenges presented by specific initiating events (e.g., Transient with Loss of PCS, Transient with PCS Available, LOSP) or document how those challenges are addressed in the top logic model.

2) Modify the existing event sequence and/or linked fault tree to ensure that the challenge to the Pressurizer Safeties is captured for initiating events that would prevent the PORVs from opening.

3) Explicitly model the SBO sequences to ensure that the necessary mitigating systems are addressed following power recovery.

Response

GTRAN was restructured to address this comment. The tree was updated to explicitly ask demand for PORVs and Safeties Page E3-9 of 87

F&O Number F&O Details 1-19 It was noted that HFE HAPRZ (discussed in Section 6.8 and Section 7.2) is not calculated using HRA Calculator. This event seems to have been carried over from the Watts Bar analysis and is treated as basic event U1_L2_NOTRCSDEPNOSBO.

In addition, although Section 6.8 says that the No RCS Dep branch is set to a value of 1 for SBO cases, the value of basic event U1_L2_NOTRCSDEPSBO in the provided MASTERL2.CAF fault tree was set to 0.9995. This also appears to be a carryover from Watts Bar.

(This F&O originated from SR LE-C7)

Associated SR(s)

LE-C7 Basis for Significance The HFEs for intentional depressurization needs to be evaluated to determine their applicability to SQN.

Possible Resolution

1) Include HAPRZ in the HRA analysis for SQN or justify the applicability of the Watts Bar value and provide an appropriate reference to the source.

2) Verify that the proper value of basic event U1_L2_NOTRCSDEPSBO is being used in the quantification.

Response

The current analysis has been updated to change the value of failure to depressurize the RCS during SBO scenarios to 1.0 (assumed failure) in the model. The basic event HAPRZ, which represents failure to depressurize for non-SBO scenarios, uses a value of 0.1 for failure to depressurize, which was taken from WCAP 16341-P, revision

0. The level 2 event trees also uses the compliment to this action called HAPRZ-SUC which has a probability of 0.9.

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F&O Number F&O Details 2-1 Section 7.0 of the Initiating Events Analysis observes a decreasing trend in initiator frequency in the more recent generic data sources. However, there is no comparison of the SQN results against the generic results nor an explanation of any significant differences.

(This F&O originated from SR IE-C12)

Associated SR(s)

IE-C12 Basis for Significance The current evaluation of the initiator frequency results does not compare SQN results to the generic frequency results.

Possible Resolution The section 7.0 discussion could be expanded to include a comparison of the SQN results to the generic results and an explanation of any significant differences.

Response

Added text to Initiating events notebook that compares Sequoyah initiator frequencies to generic industry data.

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F&O Number F&O Details 2-3 Section 4.3.1 of the Data Analysis notebook discusses the basic event probability model methodology. Generic data sources selected for use are applicable for SQN.

For those components which had a failure during the analysis time period (1/1/03 -

11/30/09), the distributions are updated via the Bayesian update program built into CAFTA program. However, the intent of this supporting requirement is to assure realistic parameter estimates are calculated for SIGNIFICANT basic events based on relevant generic and plant-specific evidence, not just those for which failures have occurred. Where no failures have occurred, use of the generic data may be conservative since it includes failures from potentially less reliable components across the industry.

(This F&O originated from SR DA-D1)

Associated SR(s)

DA-D1 DA-D3 Basis for Significance Using potentially conservative failure rates for significant components can skew the risk results. Both generic and plant-specific experience should be considered for the significant basic events.

Possible Resolution Consider performing a Bayesian update for all significant basic events - not just those for which failures have occurred.

Response

Significant contributors that were not Bayesian updated were identified as:

BATFR - Battery Fails to Operate BUSFR - Bus Fails to Operate CBKFO - Circuit Breaker Fails to Open FNSFD - Standby fan fails to start HXRPL - Heat Exchanger (River Water) Plugs or Fouls MOCXC - Motor Operated Valve Transfers Closed POEFR - ERCW pumps fail to run PSRFR - RHR pumps fail to run STRPL - Strainers plug TSCPL - Traveling water screens plug XRFR - Transformer fails to operate These events were Bayesian updated using plant specific data. The notebook has Page E3-12 of 87

been updated to reflect these additional updates.

Page E3-13 of 87

F&O Number F&O Details 2-4 Appendix F of the Data Analysis notebook provides graphs that show the prior and posterior distributions. Table 19 lists generic and Bayesian-updated mean values, along with a ratio of the posterior to prior mean value. However, there are no conclusions drawn about whether or not the posterior distributions are reasonable given the relative weight of evidence provided by the prior and the plant-specific data. (Note: the statement that "There are no significant differences between the industry data from NUREG/CR-6928 and the posterior distributions for the SQN failure rates" in section 11.0 is not judged to be sufficient. For example, the ratio of the posterior to prior mean for the AHUFR type code in Table 19 is 10.6. For type code LSTFR, the ratio is 4.3. The significance of these differences should be discussed.)

(This F&O originated from SR DA-D4)

Associated SR(s)

DA-D4 DA-E2 Basis for Significance The reasonableness check needs to assess whether the Bayesian updates yield expected results given the relative weight of evidence provided by the prior and the plant-specific data.

Possible Resolution Discuss the observed differences in the prior and posterior distributions and draw conclusions on the significance associated with those differences.

Response

The posterior distributions were validated using the following process. Using a Monte Carlo simulation, the posterior distributions were samples to see the probability of having a recurrence in the number of events observed in the data window given the number of successes in the data window. If the mean value was within 0.05 to 0.95 the resultant distribution was used within the model.

Appendix F was re-written to address this analysis as well as to present the prior, posterior, and plant specific distributions.

Page E3-14 of 87

F&O Number F&O Details 2-5 The method from NUREG/CR-6823 is used to Bayesian-update a Jeffreys noninformative prior distribution with plant-specific experience. However, there is no comparison of the posterior means to plant-specific means. (See the last sentence in NUREG/CR-6823, section 6.7.1.2.)

(This F&O originated from SR DA-D4)

Associated SR(s)

DA-D4 DA-E2 Basis for Significance A reasonableness check should be performed to assure the Bayesian-updated maintenance unavailabilites yield expected results when compared to plant-specific mean values given the amount of plant-specific data.

Possible Resolution Compare the Bayesian-updated maintenance unavailabilites to plant-specific mean values, discuss the observed differences and draw conclusions on the significance associated with those differences.

Response

The fundamental assumption used in the Bayesian update process described in the Data Analysis notebook for unavailability calculations is that there is no prior information from which to Bayesian update. Therefore, the methodology used was to use a Jefferys non-informative prior (0.5) as the foundation for the update process. All of the available data that was used was from plant specific data collection, therefore the posterior mean and plant specific mean are directly correlated. The following assumption was added to Section 3.0 to address the non-informative prior.

"For unavailability calculations, a Jefferys non-informative prior was used as there was no informative prior information available."

Page E3-15 of 87

F&O Number F&O Details 2-8 The importance of components and basic events are identified in sections 5.1 and 5.7 of the Accident Sequence notebook, respectively. However, documentation that determined the importance results make logical sense could not be identified.

(This F&O originated from SR QU-D7)

Associated SR(s)

QU-D7 Basis for Significance Multiple reviews of the model solution results yielded model changes, as documented in Table 7.0-1 and Appendix F of the Quantification notebook.

Importance measures are calculated in section 5.7; however, these need to be evaluated in light of the model solution results. In other words, do the importance measure reports yield the expected results?

Possible Resolution Document an evaluation of the importance measure results in light of the CDF results.

Response

A review of the importance of components and basic events has been performed to determine that they make logical sense. The review shows that the risk significant components are consistent with the model results and limitations. Significant contributors include basic events associated with diesels, ERCW, Component Cooling, RHR, Atmospheric Relief Valves (ARVs) and Air Compressors. In SQN, failure of the auxiliary control air headers impacts the ARVs that are needed to cooldown/depressurize in LOCA scenarios since the condenser is unavailable from a Phase B isolation. The emergency diesel, ERCW, RCP breakers, and RHR are important since their failure result in scenarios involving SBO and RCP seal LOCAs.

Page E3-16 of 87

F&O Number F&O Details 3-1 Section 4.5, The calculation above provides that the containment hole size must lie between a 1 inch equivalent path and a 4 inch path. Therefore, it is acceptable to use the NRC value of 2 inches. Based on the statement, the 1 equivalent hole should have been considered.

(This F&O originated from SR LE-D7)

Associated SR(s)

LE-D7 Basis for Significance It is unknown what the applicable break size is between 1" and 4", therefore the conservative approach is to use 1".

Possible Resolution Perform detailed analysis to ensure the use of the 2" equivalent hole is allowable or use 1" and include the additional penetrations in the containment isolation analysis.

Response

Section 4.4 discusses the reasoning for concluding that the 2 hole size is acceptable for use in the Sequoyah level 2 analysis. The reference shows that the release rate corresponding to a 1771 scfm rate would be represented by a vent line diameter greater than 1 and slightly less than 2. Because the point corresponding to 1771 scfm at 19 psig (which is half of the assumed severe containment challenge pressure) is only slightly below the 2 contour line shown in Reference 33, and there is conservatism built into both the assumed containment failure pressure and the assumed leak rates at that pressure, it is judged appropriate to use 2 as the bounding value for a large leak rate.

Page E3-17 of 87

F&O Number F&O Details 3-7 Several areas were identified that need additional discussion with respect to the Success Criteria Analysis. For example:

1) The differences between plant response to a pipe-break SLOCA and a consequential PORV LOCA are not fully discussed. Given the differences in break location, there should be some discussion in the Success Criteria Notebook of why the pipe-break SLOCA analyses bound the consequential PORV LOCA. In addition, while there is a discussion in the TH Notebook comparing the values of some key parameters for the pipe-break SLOCA and the consequential PORV LOCA, this does not fully explore differences in plant response that may affect the success criteria.

2) There needs to be more discussion of why the 480 gpm per pump RCP Seal leaks are included in the Medium LOCA (MLOCA) grouping. It is stated in Section 4.4.10 of the TH Notebook that the 480 gpm seal LOCA meets the MLOCA requirement of not requiring AFW for accident mitigation, but there is no documentation of success criteria analyses that support this statement.

3) The basis for assuming a SGTR flow of 700 gpm in Section 7.2.10 of the TH Notebook needs to be discussed in more detail than simply noting that no historic SGTR has been of the magnitude of a double-ended guillotine rupture of a SG tube.

4) The LOCA analysis is limited to the upper and lower end of the break range for each class. TH analysis at the middle of the break range within the Large, Medium, and Small LOCA categories may provide insights that have not been revealed by the upper and lower end of the break. For instance, it is not clear if sequence MLOCA-011 can be a success path for a break in the 3 to 5 inch range.

(This F&O originated from SR SC-B3)

Associated SR(s)

SC-B3 Basis for Significance There is a lack of discussion regarding how these items were treated in the success criteria analysis.

Possible Resolution Expand the discussion of the noted items in the Success Criteria documentation.

Response

1) The small LOCA events assume that he break occurs low within the physical structure of the RCS. These breaks will always have a higher deltaP value than those of breaks at the top of the RCS (PORV LOCA). Due to the additional pressure and other thermo-hydraulic characteristics the success criteria is bounding for the SLOCA cases.

2) The 480 gpm seal LOCA is now grouped as a SLOCA. This requires the use of Page E3-18 of 87

AFW for successful accident mitigation.

3) The value of 700 gpm was used as an attempt to bound the analysis. The selection of 700 gpm was done to assure that the analysis was realistic in nature, but conservative as well.

4) The MLOCA event tree has been restructured to require successful injection of the CLAs this is to assure that any break size within the MLOCA range can be successfully mitigated after failure of the CVCS system to inject.

Page E3-19 of 87

F&O Number F&O Details 3-9 All mitigation strategies credited in the accident sequence model when the high pressure recirculation has failed are not prescribed by the corresponding EOPs. In other words, the mitigation credit in the event tree model has no basis. This issue has been self-identified by the SQN PRA staff and a corrective action report has been written for the EOP group to resolve this issue. At this stage the PRA group "firmly" believes that the EOP will be modified, not the model. Thus it is a tracking issue.

(This F&O originated from SR SC-A3)

Associated SR(s)

SC-A3 Basis for Significance There is a CR written by the TVA.

Possible Resolution Tack the CR and ensure that the procedures are modified or that the model is changed to reflect the as-operated plant.

Response

EOP revisions were approved at the SQN PORC meeting on May 6th 2011.

Page E3-20 of 87

F&O Number F&O Details 3-13 Section 4.4.2 of the TH Notebook (MDN-000-000-2010-205) discusses the use of MAAP for LLOCA in the cold leg. The conclusion is that the large LOCA (LLOCA) limitations are not applicable to break sizes < 10 inches. The reference used for this is a MAAP training lecture. Use of MAAP to model the injection phase of the LLOCA needs additional justification with reference to the applicable technical documents.

(This F&O originated from SR SC-B4)

Associated SR(s)

SC-B4 Basis for Significance MAAP is known to have difficulty modeling the initial phase of the LLOCA events.

Possible Resolution Use RELAP or other alternative codes to analyze the initial phase of the LLOCA or provide a more comprehensive justification for the use of MAAP which includes benchmarking against other codes.

Response

The limitation noted for MAAP are for the larger end of the LLOCA spectrum per EPRI TR-1020236. The success criteria for the large LOCA was consistent with and largely derived from the SQN design basis analysis and SAR. While this does lead to conservative results in the LLOCA event tree, the expenditure of additional resources for the further refinement using additional codes such as RELAP is not warranted, given that LLOCA events are not risk significant in the SQN model. The low importance of the LLOCA sequences is consistent with other PWRs in the industry. The MAAP analysis for the LLOCA events were used mostly as confirmation of the event trees based on the SQN SAR and for timing of HRA events. Specifically for the HRA events, MAAP was only used to determine depletion of the RWST and long term time to core damage based on failure of hot leg recirculation. Both of these cases are significantly past the initial stages of a LLOCA where MAAP is noted to lack the thermal hydraulic detail required to evaluate the initial blowdown (EPRI-TR1020236).

Page E3-21 of 87

F&O Number F&O Details 3-14 Several documentation issues were noted in the Success Criteria and TH Notebooks.

Specifically,

1) Figures 7-60 and 7-61 of the TH Notebook (MDN-000-000-2010-205) need to be replaced with updated results.

2) The discussion of accident sequence node LPH in Section 7.3.1 of the TH Notebook (MDN-000-000-2010-205) states that The time for switchover to hot leg recirculation is specified in the EOP E-1 as 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the initiation of a large LOCA (Reference 4, Step 31c). In the paragraph immediately below this statement, the calculation of the time available for recovery from a failure of recirculation uses a switchover time of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Discussion with TVA personnel indicated that the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> value was copied from the WBN notebook. The actual time specified in the SQN procedures is 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

3) Table 7-13 of the TH Notebook (MDN-000-000-2010-205) does not include success path ISLM-014 as shown in Figure 6.4-10 of the Accident Sequence Notebook (MDN-000-000-2010-0201). In addition, success path ISLM-017 in Table 7-13 of the TH Notebook is not shown in Figure 6.4-10 of the Accident Sequence Notebook.

4) Section 4.4.11 of the TH Notebook (MDN-000-000-2010-205) discusses the classification of a Stuck Open PORV as a small LOCA. The basis needs to be provided.

(This F&O originated from SR SC-C1)

Associated SR(s)

SC-C1 Basis for Significance The documentation needs to match the current analysis.

Possible Resolution

1) Replace figures 7-60 and 7-61 with the correct figures.

2) Revise the text to use the correct information for SQN.

3) Ensure the sequence designations in Table 7-13 of the TH Notebook (MDN-000-000-2010-205) match those in Figure 6.4-10 of the Accident Sequence Notebook (MDN-000-000-2010-0201).

4) Justify the classification of the Stuck open PORV as a SLOCA.

Response

1) Figures 7-60 and 7-61 were revised in the TH calculation MDN-000-000-2010-205. In the original MAAP runs, the SG ARVs were opened at 30 minutes, this dropped pressure in the RCS. Opening of the SG ARV was not credited in the event tree for the sequences evaluated in figures 7-60 and 7-61. This is applicable to the WBN TH analysis as well.

2) The TH Notebook was revised to be consistent with EOI E-1 step 22. The correct Page E3-22 of 87

time of switching over to Hot Leg Recirculation of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> was included in Section 7.3.1 of the TH Notebook.

3) Table 7-13 and Figures 6.4-10 were revised to be consistent.

4) Additional information was included in section 4.4.11 of the TH notebook to justify the classification of a Stuck Open PORV. This information includes a comparison of core damage timing and mass/energy release rates through a SOPORV and SLOCA.

Page E3-23 of 87

F&O Number F&O Details 3-19 Section 7.2 of the HRA Notebook (MDN-000-000-2010-0204) does not explicitly discuss how the required and available manpower is addressed in the analysis.

Manpower requirements are included in the operator interview checklist as item 37.

However, it is not clear how this information was used in the development of the HEPs since some instances were observed where the operator interview responses were not used in the HRA calculator (see HFE HARR1).

(This F&O originated from SR HR-H2)

Associated SR(s)

HR-H2 Basis for Significance Some accident scenarios can require more manpower than others and this is not discussed.

Possible Resolution Add a discussion of how the manpower requirements are accounted for in the HRA, especially for those HFEs which require local actions.

Response

A discussion of the required and available manpower to perform the actions and equipment manipulations was documented in sections 7.1 and 7.2 of the HRA notebook. Also, HARR1 was revised to match the operator interview for the manpower requirements.

Page E3-24 of 87

F&O Number F&O Details 3-20 Several issues related to the TH analyses used to support the HRA were identified.

Specifically,

1) Some time windows are buried in MAAP output files which are not included in the TH Notebook and take time to review. For example, the time window for AFWOP5 is not easily available.

2) TH Notebook MDN-000-000-2010-205 Section 7.3.3 discusses the actions required following a failure of high pressure recirculation. The required action related to failure of the automatic recirculation alignment (HARR1) has two big pieces. The first is to stop the pump to avoid pump damage. If the pumps are damaged, high pressure recirculation can't be successful. The time window is short for this action and is related to RWST depletion. If the pumps are stopped on time the next action is to manually establish recirculation. The time window for that action is based on the RCS inventory depletion which is, relatively speaking, much longer.

If HP recirculation is not successful, the RCS is depressurized to facilitate low pressure recirculation (AFWOP3). These two actions (HP recirculation and RCS depressurization and establish LP injection/recirculation) are for the same mitigation function. Therefore, it is unclear why there are big differences between the time windows for these two actions. In addition, the HRA Calculator input for these actions appears to be different from the descriptions in Section 7.3.3 of MDN-000-000-2010-205.

3) The use of bounding analyses for the HFEs results in non sequence specific timing information in the HRA. For example, HARR1 is used in the accident sequences after AFWS success in SSBO and SSBI accident sequences.

However, the timing window of HARR1 is based on the medium LOCA and it is conservative for these sequences.

(This F&O originated from SR HR-I1)

Associated SR(s)

HR-I1 Basis for Significance The documentation of the TH analyses performed to support the HRA is difficult to trace and in some cases contains conflicting information.

Possible Resolution

1) Revise Table 8-1 in the TH Notebook (MDN-000-000-2010-205) to include additional information such as the time to core damage and a reference to the applicable TH cases.

2) Review the results of TH cases supporting the HRA to ensure reasonable consistency of time windows for different actions with the same purpose.

3) Refine the timing analysis as necessary to ensure the results are realistic and represent the accident sequence(s) in which the actions are used.

Page E3-25 of 87

Response

1) TH notebook revised - all HRA timing in Table 8.1

2) All TH result cases were reviewed to ensure that the time windows in use were consistent between different actions with the same purposes.

3) As stated in the details of the F&O, the analysis used is conservative. The timing analysis is for the most time limiting break for which the action is applied. This conservative timing selection addresses all potential scenarios/break sizes and would only reduce HEP and add additional margin to the analysis. This is considered to be appropriate due to the ranges of break sizes included in the broad bands of initiating event groupings. Evaluation of the recovery of additional margin from developing lower HEP individual analyses for each application of HARR1 will be completed in future revisions of the SQN PRA model.

Page E3-26 of 87

F&O Number F&O Details 3-25 Several documentation issues were noted. For example:

1) Sequences ISLM-008 and ISLM-017 were deleted from the ISLOCA event tree.

However, there is no discussion of why this was done.

2) Paragraphs in section 6.4.7 need to be revised. Specifically, the first sentence in the first paragraph on page 62, starting with "If the temperature of the RCS is 557°F and dropping, the steam dumps, S/G PORVs and blowdown isolation valves are closed." needs to be finished. There is the "if" but no "then." It is also unclear how this sentence is related to the accident sequence event tree or the following statements in the paragraph related to the PORVs.

The second paragraph on page 62 has grammatical errors (e.g., the possibility of have a RCP Seal LOCA).

3) The discussion of manual control rod insertion following ATWS in section 7.9 needs to be revised to reflect the intent to remove credit for this action from the model.

(This F&O originated from SR AS-C1)

Associated SR(s)

AS-C1 Basis for Significance Inconsistencies in the documentation can affect maintenance and update of the model.

Possible Resolution

1) Add a discussion explaining why sequences ISLM-008 and ISLM-017 are not used or re-number the sequences to ensure there are no gaps in the numbering. Also, ensure all related documents (e.g., the SC and TH notebooks) are revised for consistency.

2) Review sections 6.4.7 and 7.9 and revise, as needed, to ensure that the discussion reflects the accident sequence models.

Response

1) The sequences were not re-numbered following the latest update to the event trees. The numbering scheme will be updated in the next revision of the notebook.

2) The grammatically errors noted have been updated and revised.

3) The ATWS discussion of MRI has been updated to state that only the mechanical binding of the control rods or the failure of the automatic control system are modeled.

Page E3-27 of 87

F&O Number F&O Details 4-3 Non-water flood sources are excluded on the basis of Assumption 11 of the notebook. However, the Standard states (in Note 1 for this SR) that non-water sources should be considered, A more detailed basis for excluding these sources should be developed to meet the requirements of this SR.

(This F&O originated from SR IFSO-A1)

Associated SR(s)

IFSO-A1 Basis for Significance This is considered to be a finding since the requirements of the SR have not been fully met.

Possible Resolution Update the analysis to consider non-water sources, or better justify why the flooding impacts of these non-water sources are not significant and hence do not require evaluation.

Response

Assumption 11 was reworded to:

All sources of fluid within the plant were analyzed for flooding considerations. However, the glycol system is the only system which could have an impact on the flooding analysis. All other sources such a resin did not have enough volume to cause impact to plant operation. The glycol system also has a minimum volume, but the location of the piping, in the control rod drive rooms, causes system to be a source of spray initiating events.

Page E3-28 of 87

F&O Number F&O Details 4-7 No discussion of sources of uncertainty associated with the flooding initiating events is currently provided in the flooding notebook (MDN-000-000-2010-0203). It is noted that the notebook includes documentation of sources of uncertainty for other portions of the flooding analysis. Sources of model uncertainty for internal flooding are also documented in MDN-000-000-2010-0209, Uncertainty and Sensitivity Analysis; however, again flood initiator uncertainties are not discussed. If no uncertainties are identified for the flood initiator frequency evaluation, then the notebook should state this to be consistent with the approach used for the IFPP, IPSO, and IFSN tasks.

(This F&O originated from SR IFEV-B3)

Associated SR(s)

IFEV-B3 Basis for Significance This is considered to be a finding since the requirements of this SR are not met Possible Resolution Provide an assessment of sources of modeling uncertainty for the flood initiator frequency determination.

Response

Section 8.8 was added to the Internal Flooding Notebook with the following:

The internal flooding frequency calculation has several different uncertainties associated with the calculation. The current model uses a summation of three different frequencies, passive pipe break failures, human induced floods, and maintenance induced flooding. Each of these flooding events has its own inherent uncertainties.

For passive pipe break failures rates have been given an uncertainty parameter as presented in Section 8.5. The impact of these uncertainties can be treated by the use of a random sampling Monte Carlo process as discussed in Section 10.1.

Human induced flooding events present another difficult challenge. The use of the HRA Calculator program from Scientech creates an assumed uncertainty term for any HRA action. Since the human induced flooding events is a combination of both pre-initiating event and post initiating event, each portion has an independent uncertainty term. The HRA Calculation program also arbitrarily assigns an uncertainty term to HRA actions based on the calculated probabilities, see the HRA Calculation for more information on the uncertainty parameters (Reference 68). The other fundamental issue that is presented in human induced flooding events is the location of work. Depending on where the actual work is being performed in a flood area, isolation could be a concern as the next available valve could be in an inaccessible area. Additionally, there are no detailed procedures to address having a flood occur during a maintenance event.

Maintenance induced flooding events also present a level of uncertainty. The three Page E3-29 of 87

main inputs to the calculation of this frequency, failure rate of an MOV, mission time, and frequency of the activity all introduce some level of uncertainty into the calculation. The large internal rupture of an MOV is assumed in NUREG/CR-6928 to be a factor of 0.02 less than that of a small internal leak on an MOV (Reference 104),

as there has been no actual large internal rupture events in the industry. The mission time is also assumed based on a seven day repair interval, this number could potentially be greater than that if the component is not covered by an Technical Specification or, more likely, less than the assumed seven day repair time. The final area of uncertainty is the frequency of the activity. Most of the procedures reviewed in Appendix J have frequencies as well as conditions. These conditions could cause the actual maintenance activity to occur more times than the frequency noted in the procedure.

Page E3-30 of 87

F&O Number F&O Details 4-11 While the PRA model considers the possibility of two PORVs being blocked at the same time, there does not appear to have been an investigation of whether coincident maintenance can occur in the various SQN systems (or if coincident inter-system maintenance can occur). Therefore this SR is not met.

It was also observed that the PORV blocking basis events noted above did not appear to be documented in either the data notebook or the appropriate system notebook.

(This F&O originated from SR DA-C14)

Associated SR(s)

DA-C14 Basis for Significance This is a finding since the technical requirements of the SR are not met.

Possible Resolution A study should be conducted to determine if coincident maintenance conditions can occur. If so, the system models may need to be modified and additional basic events to represent the coincident maintenance states would need to be added. If it is determined that no coincident maintenance can occur, then this should be documented in the data notebook or within the system notebooks.

Documentation for the calculation of the time that either one or both PORVs can be blocked should also be added to either the system notebook or the data notebook.

Response

The following was added to the data analysis notebook to address coincident maintenance:

Coincident maintenance is scheduling maintenance where multiple SSCs are out of service at the same time. Specifically components on the same train, RHR train A and SI train A for example, being out of service for maintenance at the same time. The Outage and Site Scheduling Directive Manual 1.0 (Reference 28) dictates that:

Twelve (12) week schedule by FEG groups ensures that within a train week, no two (2) accident mitigating devices are removed from service at the same time [i.e., A train Residual Heat Removal (RHR) is not removed from service at the same time as A train Containment Spray.]

This requirement is further discussed in the Outage and Site Scheduling Directive Manual 4.7 (Reference 29) which states that any systems important to PRA that are unavailable at the same time must meet the requires of the plant risk matrix.

Normally maintenance on any systems important to the PRA is not scheduled at the same time. If it is these instances are extremely rare and the current model does not exclude coincident maintenance events from appearing in a single cutset. Therefore the probability of having coincident maintenance events is extremely rare and accounted for during the normal cutset processing.

Page E3-31 of 87

Page E3-32 of 87 F&O Number F&O Details 5-2 Some HFEs are set to a value of 0.0 for quantification. For example, HACI1 and HAAE1 are recovery actions for automatic signals ANDed with the signal logic.

However, the HRA analysis sets the HEP probability to 0.0 based on an analysis that the operator action is not required. This screening approach, combined with the model structure, removes the auto actuation contribution to mitigating system failure during quantification.

(This F&O originated from SR HR-G1)

Associated SR(s)

HR-G1 Basis for Significance Screening HFEs using a value of 0.0 remove the auto actuation hardware failure contributions in the quantification results.

Possible Resolution Revise the model to remove non-credited operator recovery actions from the linked fault tree or set all non-credited events to TRUE during quantification.

Response

For those events where 0.0s were used in the model the fault tree was updated to remove the events so that the conflict concerning an AND gate and a zero event will not longer be encountered during normal quantification.

Page E3-33 of 87

F&O Number F&O Details 6-2 The justification for excluding plant data prior to July 2002 in the calculation of plant specific IE frequencies is not documented well enough to support IE-C2.

(This F&O originated from SR IE-D2)

Associated SR(s)

IE-C2 IE-D2 Basis for Significance A justification was provided during the review, but it is not documented in the notebook.

Possible Resolution Include a justification for excluding data prior to July 2002 in the IE notebook.

Response

Added discussion to notebook stating that date range was adequate to get a good sample of plant data without going too far back and including events that occurred when the plant may have had different procedures and operating practices Page E3-34 of 87

F&O Number F&O Details 6-3 The alignment flags in the ERCW system are not fully implemented to represent the system alignment within the Initiating event portion of the tree. For example, the gates under U0_AEX_G006 should contain flags to indicate which pump is running and which two pumps are not, so that the two non running pumps would have considerations for failures to start.

(This F&O originated from SR IE-A6)

Associated SR(s)

IE-A6 IE-C10 Basis for Significance The current fault tree configuration does not properly account for the system alignment.

Possible Resolution Include the alignment flags in the indicated and similar gates. Review the remainder of the tree to ensure that the alignments are properly identified.

Response

The current flag alignment for ERCW has been revised so, for the baseline model, without setting a specific configuration, the flag files were set to the respective time in each configuration to that a probability is now used not a true or false value.

Page E3-35 of 87

F&O Number F&O Details 6-5 The support system initiating event trees for the most part include provisions for common cause failures and routine system alignments. There are some discrepancies in the modeling of common cause failures in the ERCW and CCS models that require attention, however. For example:

1) While a common cause event for all 3 of the 1A, 1B, and C-S pumps failing to run exists, there are not events for the 1A and C-S pumps or the 1B and C-S pumps.

2) The structure of the ERCW tree is such that pump common cause failures could result in a pump failing due to an independent failure as well as a common cause failure in a single cutset. (See gate U0_AEX_G001)

3) The common cause initiating event group U0_ERW08POEFRI is not valid, since it is entirely based on 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> exposure time for all the components. The common cause failure frequencies are therefore overestimated. The CCS tree uses a different approach than the ERCW tree for common cause initiating events. An alternate approach is also given in EPRI reports 1013490 and 1016741.

(This F&O originated from SR IE-A6)

Associated SR(s)

IE-A6 IE-C10 SY-B3 Basis for Significance Support system initiating event failures are inconsistently applied among the support system models, and may be giving incorrect results.

Possible Resolution Review initiating event common cause events and select a consistent modeling approach among the support system initiating event models.

Response

With respect to the common cause failure of the CCS pumps:

The common cause failure of the 1A and the C-S pump or the 1B and the C-S pump would not meet the requirements to cause an initiating event for the CCS system. Only failure of the A train would cause the plant to have to trip as the loads on the common train are not required for operation at power. Therefore only the common cause failure of all three pumps is modeled in the fault tree.

With respect to the common cause failure events from the ERCW fault tree:

The common cause failure events in the ERCW system where common cause failure and independent failures show up in the same cutset present a minimal and conservative impact.

With respect the common cause calculation of basic events:

Page E3-36 of 87

The common cause failure rates for ERCW pumps failing to run and CCS pumps failing to run were revised based on the EPRI document 1013490 using the discussion presented on page 5-8. The assumptions and calculation of these basic events is noted in Appendix B of each calculation.

Page E3-37 of 87

F&O Number F&O Details 6-6 Section 5 of the IE notebook shows a Bayesian process was used to combine plant specific and generic data. However, LOCA frequencies from NUREG-1829 were also updated with plant specific data. Since the frequencies in NUREG-1829 were based on expert judgment and not actual industry data, and it is not expected that a plant would experience such an event, it does not seem appropriate to use the Bayesian update process for these events. The update did not appear to significantly alter the IE frequencies, however, so there is little impact on CDF.

(This F&O originated from SR IE-C4)

Associated SR(s)

IE-C4 Basis for Significance Frequencies in NUREG-1829 were based on expert judgment and not actual industry data, and it is not expected that a plant would experience such an event, it does not seem appropriate to use the Bayesian update process for these events.

Possible Resolution Use the frequencies derived from NUREG-1829 without Bayesian updating with plant data.

Response

The frequencies presented in NUREG-1829 represent the best estimates available at that time. There is no restriction on updating an expert solicitation, as the update process will only serve as to better estimate the actual failure rate for the initialing events.

Page E3-38 of 87

F&O Number F&O Details 6-7 Section 6 of the Initiating Events Analysis, the associated system notebooks, and the HRA notebook document the use of plant-specific information in the assessment and quantification of recovery actions where available, in a manner consistent with the applicable HR SRs.

An issue was noted with the ERCW initiating event tree. Event HAAEIE "Operator Fails to Start ERCW Pump (Initiating Event)" has been set to zero based on an analysis that found one pump was sufficient to cool plant loads, so if one of the two running pumps trips, operator action is not required to start another pump. Operator action to start a standby pump would be required, however, if flow was to be lost from both running pumps. The current model essentially assumes a successful operator action to start both of those pumps.

(This F&O originated from SR IE-C11)

Associated SR(s)

IE-C11 Basis for Significance The operator action HAAEIE has inappropriately been assumed to be 100%

successful.

Possible Resolution Re-evaluate the failure rate for operator action HAAEIE, given the revised requirements of the ERCW system with regards to causing an initiating event.

Response

The ERCW initiating event model has been updated.

Calculation CN-NUC-SQN-MEB-MDQ-000-067-2000-0095 revised the existing success criteria used in the initiating event model. The results of the calculation that as long as the containment spray heat exchangers were not in service, the maximum required flow on the ERCW system would be roughly 9,000 gallons. This is within the design flow rate of 10,000 gallons per minute from one ERCW pump. Due to the change in the success criteria, the initiating event model was update to requiring the failure of two running ERCW pumps as well as failure of both standby ERCW pumps to start. The HRA action HAAEIE was added to the model under the appropriate failure to start gate, no longer under an AND gate.

Additionally, the fault tree logic in question was update so that failure to start takes into account the failure of operation action HAAEIE.

F&O Number F&O Details 6-10 Tables 42 and 43 of MDN-000-000-2010-0209 contain a list of modeling assumptions and their impact on the PRA model. However, the majority of items in Table 43 have an impact of Unknown. Classification of model impact for these assumptions is Page E3-39 of 87

necessary to meet this SR.

(This F&O originated from SR QU-E4)

Associated SR(s)

QU-E4 QU-F4 Basis for Significance The SR requires identification of the impact of identified assumptions on the model.

Possible Resolution Provide an evaluation of an impact of the items listed as Unknown in table 43.

Response

The Uncertainty and Sensitivity Analysis calculation has been updated in the following ways:

Text concerning the discussion of Unknown impacts and performing a respective uncertainty analysis was removed from Section 5.0.

Table 43 was updated to remove the column "Model Impact" and the column "Comments" was updated to "Model Impacts and Comments" and expanded.

Page E3-40 of 87

F&O Number F&O Details 6-12 From the results presented in sections 5.2 and 5.7 of MDN-000-000-2010-0208, it can be inferred that the definition of significant basic event and significant accident sequence are consistent with those listed in Part 2 of the standard. This is not explicitly stated in the documentation, however. The definition of significant cutset is not provided, nor does the 100 cutset list provided in the documentation imply that the part 2 definition was used, as the 100 cutsets do not represent 95% of the risk.

(This F&O originated from SR QU-F6)

Associated SR(s)

QU-F6 Basis for Significance Documentation of the definition of "significant" is required by the SR.

Possible Resolution Provide a definition of significant cutset, significant sequence, and significant basic event in the documentation.

Response

The documented definition in Section 1-2.2 of the ASME/ANS combined standard was added to the quantification calculation.

Page E3-41 of 87

Suggestion Level F&Os F&O Number F&O Details 1-1 RG 1.200 Revision 2 documents a qualified acceptance of this SR. The NRC resolution states that to meet Capability Category II, the impacts of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement) must be qualitatively assessed using conservative assumptions.

(This F&O originated from SR IFSN-A6)

Associated SR(s)

IFSN-A6 Basis for Significance This is an enhancement required to satisfy the RG 1.200 qualification.

Possible Resolution Document a qualitative assessment of the impacts of jet impingement, pipe whip, humidity, condensation, temperature and other flood-induced mechanisms that are not explicitly modeled.

Response

The analysis was changed so that all components within a flood area are failed on initiation regardless of the equipment qualifications or other HELB mitigation features.

Page E3-42 of 87

F&O Number F&O Details 1-2 All components were assumed failed if subjected to submergence. MDN-000-000-2010-0203 Appendix F documents whether components were spray vulnerable.

Discussion with the responsible analyst revealed that factors considered in spray vulnerability determinations included shielding, sealing, and equipment qualification records. However, this is not documented in the notebook.

(This F&O originated from SR IFSN-A7)

Associated SR(s)

IFSN-A7 IFSN-B2 Basis for Significance The technical requirement is met, but the documentation could be enhanced.

Possible Resolution Document the basis for determining whether a component is "spray vulnerable" in MDN-000-000-2010-0203.

Response

The following text was added to Appendix F:

During the walkdowns spray vulnerability was determined by observations of the components. For MOVs, an obvious seal with water proofing had to be observed to determine if the component was vulnerable to spray.

After the walkdowns were complete a comparison to the EQ database was done to observe if any components which were seen during the walkdowns and noted as being vulnerable to spray actual were environmentally qualified. Those components were removed from the spray analysis as well as Appendix G and Appendix H.

Tables were also added to Appendix F to show those components that are currently listed as environmentally qualified in the MAXIMO database.

Page E3-43 of 87

F&O Number F&O Details 1-3 The water depth required to cause failure of doors is documented in the Appendix F walkdown sheets, but the derivation of this value is not documented.

(This F&O originated from SR IFSN-A9)

Associated SR(s)

IFSN-A9 IFSN-B2 Basis for Significance Discussion with the responsible analyst revealed that holdup of water by doors was not credited in the final analysis. Therefore this is a documentation issue that does not affect the results.

Possible Resolution Document the basis for the water depth required to cause door failure used in Appendix F.

Response

Door failure height calculations were performed post walkdown. The failure heights were broken into two different sections. If the door was observed to be a fire door, the calculation of failure height used the HELB analysis, if the door was wire mesh or a non-fire door then the failure height was assumed to be 0 feet.

The height of water necessary to fail a door is calculated based on if the height will exceed the actual height of the door. For calculations where the height of water is less than the height of the door, the following equation can be used:

2

=

Where p is the failure pressure and is the specific weight of water.

For those calculations where the height of water would exceed the actual height of the door, a different equation must be used. The door is now considered to be a completely submerged surface, so the following equation can be used:

= +

2 Where p is the failure pressure, is the specific weight of water, and hdoor is the actual height of the door.

For the purposes of Appendix F, the failure pressure of all fire doors was taken from the HELB analysis, which states that a pressure of 1.5 psid will cause the doors to fail. Using the equations above, the failure height of water is calculated to be 6.92 feet.

F&O Number F&O Details 1-5 Documentation of the scenario impacts needs to be strengthened in some areas. For example:

Page E3-44 of 87

1) In MDN-000-000-2010-0203 Section 9.4.3.3 for flood area 734.0-A13 it is stated that "the flood frequency for those events that impact both ACAS compressors will be 1/3 of the original frequency." No basis is provided for this statement.

2) Section 7.3 makes the general statement that for the Turbine Building "flood originated in any level would propagate freely to the basement of the building without any hindrance." This same assumption was applied in partitioning Auxiliary Building area 930.0-A1, but this is not documented.

(This F&O originated from SR IFSN-A6)

Associated SR(s)

IFSN-A6 IFSN-B2 Basis for Significance The technical process is acceptable, but the documentation could be strengthened.

Possible Resolution Document assumptions pertinent to the flood scenarios such as those noted in the F&O description. This could be done in the Notebook or the FRANX database.

Response

The text quoted in the analysis has been updated. A further walkdown of the plant was done to reflect piping that could impact both the ACAS air compressors. Further walkdowns conducted in April 2011 showed there was no piping that was within twenty feet of both air compressors skids on the refuel floor. Therefore the section of the notebook addressing such issues was removed.

Additionally, the breaking apart of 690 into zones has been enhanced in the documentation and is discussed in the plant partitioning section. The documentation of the flooding scenarios and the flooding analysis now includes the partitions as part of the analysis, not as a change after review of the cutsets.

Page E3-45 of 87

F&O Number F&O Details 1-6 MDN-000-000-2010-0204 Section 7.3 states that a reasonableness check was performed and generally describes the factors considered. However, it would be helpful to provide tables that grouped HEPs by the relevant factors to support the conclusions reached.

(This F&O originated from SR HR-G6)

Associated SR(s)

HR-G6 Basis for Significance The check was performed, but the results are not documented in a way that supports verification of the conclusions.

Possible Resolution Include a table that groups the HFEs by the complexity to support the conclusion that "HFEs that are more complex have higher failure probabilities than simple actions."

Similarly, a table that groups the HEPs by the available time would support the conclusion that "HFEs with shorter time windows have higher failure probabilities due to factors including insufficient time to credit review from the STA and negative performance shaping factors (e.g., high stress levels). These factors are not included in the referenced Table 10-2.

Response

Tables have been added to HRA notebook section 7.3 for the complexity and time margin comparisons completed for the SQN Rev 5 PRA and HRA model update.

Page E3-46 of 87

F&O Number F&O Details 1-9 The CCS system success criteria is modeled as dependent on the temperature of the ERCW system. MDN-000-070-2010-0217 Appendix D documents the derivation of the probability for flag event FLG0070_ERCW_TEMP_GT_70 (FLG_0024SUMMER in the linked fault tree). However, the temperature data is presented in a graphical format rather than a tabular format. A tabular format would make it easier for a reviewer to perform a validation of the data.

(This F&O originated from SR SY-A10)

Associated SR(s)

SY-A10 Basis for Significance This is a documentation issue not affecting the technical quality of the model.

Possible Resolution Provide the data used to derive FLG0070_ERCW_TEMP_GT_70 in a tabular format.

Response

The data used to create the graph in the CCS notebook is 111,210 cells long. This data is not feasible to be presented as part of the system notebook.

An electronic copy will be kept with the notebook and will be available to any person wishing to review the data.

Page E3-47 of 87

F&O Number F&O Details 1-12 Some sources of uncertainty which are characterized as having an impact on LERF are not analyzed using sensitivity analysis (e.g., core melt arrest in vessel and modeling of the ARFs).

In addition, Section 5.14.10 states that Although post-core damage human actions such as intentional depressurization of the RCS are modeled as realistically as possible, the uncertainty related to these actions is addressed in sensitivity studies.

The notebook references the sensitivity and uncertainty analysis notebook as the location for these studies. However, MDN-000-000-2010-0209 performed the uncertainty by setting the values of HEPs in the model to their 5th and 95th percentile values. This does not address uncertainty in the value of the HFE for intentional depressurization of the RCS which is modeled as part of the U1_L2_NOTRCSDEPNOSBO basic event.

(This F&O originated from SR LE-F3)

Associated SR(s)

LE-F3 Basis for Significance This is a completeness issue which would enhance the analysis.

Possible Resolution Add sensitivity analyses to address these additional sources of uncertainty in the LERF results.

Response

The HFE for intentional depressurization of the RCS was changed from U1_L2_NOTRCSDEPNOSBO to HAPRZ in the model, and was included in the uncertainty study for the HFEs.

Page E3-48 of 87

F&O Number F&O Details 1-13 The operator actions reflected in the event trees, and the sequence specific timing and dependencies that are traceable to the HRA for these actions are discussed in the HRA notebook. The operator actions modeled for each sequence are listed as a separate subsection in MDN-000-000-2010-0201. However, it is suggested that a summary discussion of operator actions affecting the accident sequences, including a discussion of the top events impacted, be included in the AS notebook.

(This F&O originated from SR AS-A4)

Associated SR(s)

AS-A4 AS-C2 HR-G4 Basis for Significance Documentation enhancement. The links can be identified through cross reference to the system, success criteria, and HRA notebooks.

Possible Resolution Provide a cross reference for each operator action to the affected event tree top event in the Accident Sequence Notebook.

Response

The operator actions are discussed along with each of the top events for all event trees in the accident sequence notebook.

Page E3-49 of 87

F&O Number F&O Details 1-17 Calculation Type 1 is used for mission time events and Type 2 is used for basic events where the probability is based on periodic tests. The CAFTA users manual states that "it is better to use the more precise formulas of calculation types 3, 4, 5 or 6. This is especially important if you are using larger numbers (e.g., t > .05), or if you will be doing uncertainty analysis." (see CAFTA users manual Tables 6-2a and 6-2b and the text below Table 6-2b.)

(This F&O originated from SR QU-E3)

Associated SR(s)

QU-E3 Basis for Significance The current method yields a valid approximation of the basic event probabilities, but does not represent the recommended practice for CAFTA.

Possible Resolution Use the more precise formulas for the time dependent basic event probability calculations.

Response

The current model does not have any events where the lambda*t value approaches 0.05.

The models used will have their calculation types updated to 3 and 5 where appropriate when any update is to be performed.

For the current model, no events generated a random value greater than 1.0 for the uncertainty graphs.

Page E3-50 of 87

F&O Number F&O Details 1-18 The system notebooks typically state that high energy line break (HELB) is considered in the Internal Flooding Analysis (e.g., ERCW Notebook Section 3.4.7.2.6). However, assumption 3.1 of the IF Notebook states that Additional failure modes; jet impingement, pipe whip, humidity, condensation, and temperature-induced failures are outside the scope of this analysis.

(This F&O originated from SR SY-C2)

Associated SR(s)

SY-C2 Basis for Significance The treatment of HELB is not clearly documented in the system notebooks.

Possible Resolution Modify the statements in the system notebooks to clearly state that HELB is not treated and to provide a justification for this.

Response

The assumption was removed from the IF notebook. The HELB events are now addressed in the IF analysis.

Page E3-51 of 87

F&O Number F&O Details 1-20 The SQN PRA considered Early containment failure as well as Late containment failure and basemat melt through. After containment failure, there is no additional equipment nor human action credited to mitigate the consequences.

There is also no evidence that a review was performed to determine if crediting operation of additional equipment or human actions after containment failure would reduce LERF.

(This F&O originated from SR LE-C11)

Associated SR(s)

LE-C11 LE-C12 Basis for Significance Recommendation for meeting the requirements for Capability Category II/III.

Possible Resolution Document a review of the LERF results to determine if credit for equipment operation after containment failure or additional operator action credit would be effective in reducing LERF.

Response

There are no additional actions or equipment currently credited in the level 2 analysis to mitigate the consequences of a release after containment failure. This results in somewhat conservative results. A review has been performed to determine if crediting additional equipment or crediting additional human actions could result in a LERF reduction. An action identified during this review involves crediting manually closing the RCP seal water return outboard isolation valve following core damage in the event that it fails to close on demand. A sensitivity study was performed to determine the effect of this action using various assumed failure probabilities (see Section 12.7), although the feasibility of implementing the action has not been studied in detail.

Page E3-52 of 87

F&O Number F&O Details 1-21 A thorough list of references is documented with each system notebook. However, the reference revision level is not always included (see the Diesel Generator and RPS notebooks, for example.)

(This F&O originated from SR SY-A2)

Associated SR(s)

SY-A2 SY-C2 Basis for Significance The supporting requirement is met, but the documentation could be enhanced.

Possible Resolution Provide the applicable revision level for each reference to improves traceability of the source documents.

Response

Reference levels were left off of the references in all system notebooks consistent with the TVA practices for PRA calculations.

Page E3-53 of 87

F&O Number F&O Details 2-2 Per discussion with the SQN PRA data analyst, maintenance is generally performed on a train basis rather than across redundant components. Where redundant maintenance is permissible, e.g., the ERCW system, which has 8 pumps, the fault tree allows for the generation of cut sets that have multiple pumps in maintenance.

However, it would be helpful to document a verification that simultaneous unavailability of redundant equipment is not how work is planned.

(This F&O originated from SR SY-A20)

Associated SR(s)

SY-A20 SY-C2 Basis for Significance Since maintenance on redundant equipment is not modeled as a planned event, documentation should be provided or referenced that describes how maintenance is planned/coordinated. This assures that any maintenance dependencies are not overlooked.

Possible Resolution Document the maintenance approach taken on redundant equipment and any impact on the PRA model.

Response

Coincident unavailability is now discussed in Section 7.4.4 of the data analysis notebook.

Page E3-54 of 87

F&O Number F&O Details 2-6 No modifications to plant design or operating practices were identified that lead to a condition where past data are no longer representative of current performance.

Thus limiting the use of old data was not required. However, for completeness, it is suggested that the data analysis document the consideration of this supporting requirement.

(This F&O originated from SR DA-D8)

Associated SR(s)

DA-D8 DA-E2 Basis for Significance No documentation addressing this supporting requirement was identified.

Possible Resolution Document a consideration of modifications to plant design or operating practices that could lead to a condition where past data are no longer representative of current performance in the Data Analysis notebook.

Response

Section 7.2.1 was added to the data analysis notebook to address plant design changes.

Page E3-55 of 87

F&O Number F&O Details 2-7 The data analysis aligns well with the PRA Standard requirements and is generally well-documented. Adding a 'roadmap' to the PRA Standard data SRs - as was done elsewhere in the PRA documentation - would enhance the performance of PRA applications, upgrades, and peer review.

(This F&O originated from SR DA-E1)

Associated SR(s)

DA-E1 Basis for Significance Adding a 'roadmap' to the PRA Standard data SRs would enhance the performance of PRA applications, upgrades, and peer review.

Possible Resolution In the Data Analysis notebook, add a 'roadmap' to the PRA Standard data SRs.

Response

Appendix I was added to the notebook to address the ASME/ANS standard sections.

Page E3-56 of 87

F&O Number F&O Details 3-4 The success criteria description needs to include the boundary conditions such as RCS pressure. In general, it is not clear what the condition is that allows the SI pumps to operate.

(This F&O originated from SR AS-A3)

Associated SR(s)

AS-A3 AS-B2 SC-A3 Basis for Significance Documentation of required conditions permitting some equipment to operate is not provided.

Possible Resolution Provide more detailed discussion of the success criteria and mitigation system operating characteristics (e.g. pressure, flow rate) and how the conditions are achieved. For instance, the SI pump injection pressure and how the pressure is achieved in the accident sequence (i.e., by opening pressurizer or SG PORVs) should be discussed.

Response

All boundary conditions are listed in either the parameter file or the input decks electronic copies of these are available on request.

Page E3-57 of 87

F&O Number F&O Details 3-10 The EOPs associated with a specific accident sequence success path are not identified in the Thermal Hydraulic Analysis or the Success Criteria Notebook. These are also not explicitly discussed in the Accident Sequence Notebook.

(This F&O originated from SR AS-A5)

Associated SR(s)

AS-A5 Basis for Significance There is no discussion relating the Emergency Operating Procedure (EOP) with accident progression in AS notebook.

Possible Resolution Provide discussions relating EOPs to the accident sequence and top events ordering.

Response

The EOP steps are now incorporated into the accident sequence notebook.

Page E3-58 of 87

Appendix A - Resolution of F&Os Facts and Observations Summary - Suggestion F&Os F&O Number F&O Details 3-11 Use of the design basis for certain success criteria may result in conservative modeling. For example:

1) Section 4.4.3 of the TH Notebook (MDN-000-000-2010-205) discusses the MAAP 4.0.7 limitations which prevent use of MAAP for determining the number of accumulators required for Large LOCA (LLOCA) success. The use of the design basis assumption that 3 of 3 intact loop accumulators are required is likely to be conservative.

2) Section 4.4.7 of the TH Notebook (MDN-000-000-2010-205) discusses the number of lines needed for the Emergency Core Cooling System (ECCS). Based on MAAP limitations, the conclusion is that the current analyses only support ECCS flow through all intact lines. This conclusion is likely to be conservative for some sequences.

(This F&O originated from SR SC-B1)

Associated SR(s)

SC-B1 Basis for Significance The use of design basis success criteria for the accumulators and for the required number of injection paths may be conservative.

Possible Resolution Perform PRA specific analysis using an alternative code to determine if success can be achieved with fewer than 3 accumulators or with flow to fewer ECCS injection paths.

Response

MAAP currently is the consensus model of choice for analysis supporting the development of the PRA model. The use of other codes does not facilitate the development of the PRA model. The current success criteria of 3 of 3 CLAs will be retained in the model.

Page E3-59 of 87

F&O Number F&O Details 3-22 It is stated that the impacts of the initiating event on mitigation systems are captured in the top events. However, there is no discussion of these impacts in the accident sequence notebook.

(This F&O originated from SR AS-B1)

Associated SR(s)

AS-B1 Basis for Significance There is no documentation found that explicitly describes the dependencies between the mitigation systems and the initiating events.

Possible Resolution Discuss the impact of initiating events on individual mitigation systems under each top event. Alternatively, provide an initiating event to mitigating system dependency matrix.

Response

An initiating event impact table was added to the Success Criteria Notebook Page E3-60 of 87

F&O Number F&O Details 3-23 The impact of the phenomenological conditions created by the accident progression is not discussed in the accident sequence notebook.

(This F&O originated from SR AS-B3)

Associated SR(s)

AS-B3 Basis for Significance There is no discussion of phenomenological conditions in the AS notebook.

However, the environmental conditions affecting equipment operation is captured in the system analysis notebooks.

Possible Resolution Add a discussion of the phenomenological conditions created by the accident sequence and their impact on the credited mitigation equipment.

Response

The current phenomenological conditions, initiator impact, are discussed within each system notebook.

Page E3-61 of 87

F&O Number F&O Details 3-24 The intersystem dependencies are embedded in the accident sequences, but there is no explicit discussion of these dependencies.

(This F&O originated from SR AS-B5)

Associated SR(s)

AS-B5 Basis for Significance There is no intersystem dependency discussion in the AS notebook.

Possible Resolution

1) Add an explicit discussion of the intersystem dependency to the discussion of each accident sequence, or

2) Include a system dependency matrix in the AS Notebook to illustrate the dependencies.

Response

A system dependency matrix has been included within the SC notebook.

Page E3-62 of 87

F&O Number F&O Details 4-1 Section 5.2 of the Internal Flooding notebook (MDN-000-000-2010-0203) considers flood areas in the buildings of both units, and includes all common buildings. At the building level, the text discusses whether the building contains shared equipment; however, the text and tables do not indicate which specific flood areas can impact both units. It would be helpful to enhance the documentation to indicate which flood areas have multi-unit impacts.

Similarly, the discussion of food sources should attempt to identify sources with multi-unit impacts.

(This F&O originated from SR IFPP-A3)

Associated SR(s)

IFPP-A3 IFSO-A2 Basis for Significance This is a suggestion since it pertains solely to enhancement of the documentation of the flood area partitioning and flood source identification process. The flood analysis itself correctly addresses multi-unit impacts.

Possible Resolution Include (in the text of section 5.2 or within the tables of included areas) indication of what areas have multi-unit impacts. Include similar documentation in section 6.1 for flood sources.

Response

All areas currently analyzed that contain ERCW, CCS, HPFP, RCW or any other infinite source of water are addressed in Section 5.2 The tables provided list all areas of the plant including those where there are and are not multi-unit impacts.

No changes were made to the internal flooding document Section 5.2 or 6.1 Page E3-63 of 87

F&O Number F&O Details 4-6 The SQN flooding analysis has addressed some, but not all of the requirements for Category II/III. EPRI flooding data based on generic industry experience is used for flood initiating events due to pipe ruptures. Plant-specific data that might influence the pipe failure data (e.g., material condition of the fluid systems and water hammer experience) are not considered. However, a review of plant-specific maintenance-induced flooding events was performed (Appendix G of the flooding notebook) and was considered in the calculation of maintenance flooding frequency. To fully meet Category II/III, an assessment should be made of plant material condition and water hammer experience) and whether plant conditions warrant any adjustments to the generic flood frequencies that are used.

(This F&O originated from SR IFEV-A6)

Associated SR(s)

IFEV-A6 Basis for Significance This is judged to be a suggestion, since the response to this F&O will most likely only impact documentation. Also, the analysis meets the Category I requirements, which may be sufficient for most applications Possible Resolution Review plant-specific experience pertaining to plant material condition and water hammer and document the results of the review in the flooding notebook.

Response

A review of plant specific flooding events was performed in Appendix G. This data was incorporated into the analysis for initiating event frequencies.

Page E3-64 of 87

F&O Number F&O Details 4-8 Dependency between pre-initiator events was determined to not be applicable due to the large amount of time between test and maintenance events between various system trains and the use of different crews to perform each train's activities.

Dependence between pre-initiators and post-initiators is also not appropriate. The rationale for not considering dependency for pre-initiators seems appropriate.

However, several inconsistencies were noted in the HRA notebook documentation (MDN-000-000-2010-0204) concerning the pre-initiator dependency treatment.

Various HRA calculator entries for the pre-initiator events (in Appendix B) indicate that dependency between events is to be considered (see for example, event SHEEMC_4). The introductory material in Appendix F contains some statements indicating pre-initiator dependency will be considered, and other statements explaining why dependency between these events is not expected. These inconsistencies should be corrected.

(This F&O originated from SR HR-D5)

Associated SR(s)

HR-D5 Basis for Significance This is considered to be a suggestion since it pertains to correcting documentation errors. The underlying analyses themselves are correct and will not be impacted by these errors.

Possible Resolution Correct the documentation errors in the HRA notebook as noted.

Response

The documentation errors in Appendix B of the HRA notebook were corrected.

Page E3-65 of 87

F&O Number F&O Details 4-9 Appropriate generic data sources appear to be used in the SN PRA, as documented in the data analysis notebook (MDN-000-000-2010-0202). Component failure rates are taken primarily from NURG/CR-6928 (with other sources used in cases in which data for specific component types are not available). Common cause data is obtained from recent NRC (INEL) and PWROG data sources. Offsite power recovery data is obtained from NUREG/CR-6890. Component recovery is not used. Table 2 and Appendix A describe the boundaries assumed for each major component type.

The SQN PRA makes use of generic unavailability data from NUREG/CR-6928 for components for which plant-specific data is unavailable (as noted in Table 8). It is assumed (see Assumption 1 in the data notebook) that all generic data is applicable to SQN; however, since this SR requires that the consistency of the SQN practices and philosophies be checked against the generic data source assumptions, additional documentation needs to be provided to better meet the requirements of this SR. It is recognized that assumption 1 is listed as an important uncertainty and is discussed in the Uncertainties notebook(MDN-000-000-2010-0209). However, since the unique attributes concerning the use of generic unavailability data are not discussed, adding an additional assumption item for this issue may be appropriate.

(This F&O originated from SR DA-C1)

Associated SR(s)

DA-C1 DA-E2 Basis for Significance This is considered to be a suggestion as it pertains primarily to a documentation enhancement. The use of generic unavailability estimates for some plant components is probably acceptable; however documentation of the basis for accepting this data as appropriate to SQN is required.

Possible Resolution Enhance the documentation in section 6.2 to better describe the acceptability of the generic estimates for SQN. Consideration should be given to specifically identifying this generic data use as an important assumption in the Uncertainties notebook (MDN-000-000-2010-0209) as well. That notebook has an overall item concerning the use of generic data; however, a specific item for the use of generic unavailability data could also be added.

Response

A discussion of the component boundaries and maintenance practices was added to section 6.2.

F&O Number F&O Details 4-10 Failure data records are obtained from the plant's Cause Determination and Evaluation (CDE) records that are prepared by system engineers in response to Page E3-66 of 87

failure events. The guidance for CDE development in plant procedure SPP-6.6 describes bases for failures, discusses degraded conditions, and notes that Technical Specification failures or operability issues are not automatically Maintenance Rule functional failures (or PRA failures).

The CDE records are also then reviewed by the PRA staff to determine if a PRA failure has occurred. The CDEs that were used in the data analysis are included in Appendix D of the data notebook (MDN-000-000-2010-0202).

Because there are several DA-C SRs that specify requirements for the data collection and analysis process, it is suggested that the data analysis documentation be enhanced to specifically note these requirements and how they are met, especially since the other plant procedures do not specifically state these requirements (since the procedures are for system engineers and other non-PRA personnel).

(This F&O originated from SR DA-C4)

Associated SR(s)

DA-C4 DA-C5 DA-C11 DA-C12 DA-C13 DA-E2 Basis for Significance This is a suggestion since it pertains to enhancing the documentation to place all of the data analysis ground rules within the data notebook for clarity.

Possible Resolution Enhance the data analysis notebook to specifically list the data collection requirements for DA-C4, C5, C6, C11, C12, and C13.

Response

DA-C4 Functional failures are determined based on the system engineer and maintenance rule expert panel. These determinations are outlined in SPP-6.6, and are only made by a qualified individual.

DA-C5 There was an identified event where multiple repeat failures occurred within the same time. Each of these events was assigned a specific CDE, however as noted in the documentation of CDE 1615 the three events were all assigned to one failure event in the PRA model.

DA-C11 Unavailability is defined in the maintenance rule technical instruction TI-4. The definition presented states that unavailability is only counted while at power (mode 1), additionally in the definition, unavailability is credited when the component Page E3-67 of 87

would not be able to perform its designed function.

DA-C12 The definition of the component boundaries for tracking unavailability are documented in TI-4. For frontline systems only front line impacts are assigned to that system. If the ERCW header or other multi-system impact components are unavailable then the unavailability is tracked at that level.

DA-C13 For all significant unavailabilities, start and finish times are accurately documented in the maintenance rule spreadsheets.

Page E3-68 of 87

F&O Number F&O Details 4-12 Surveillance testing intervals are used for only a subset of the components for which operating data is calculated. For those using the surveillance testing intervals, the data notebook (MDN-000-000-2010-0202, section 7.3.4) does not indicate that a review of the surveillance test procedures was performed to determine if all subelements were tested on the same frequency. Therefore the Category I requirements are satisfied for this SR.

(This F&O originated from SR DA-C10)

Associated SR(s)

DA-C10 Basis for Significance This is considered a suggestion, since Category I is met and may be adequate for most applications. It is unlikely that the data estimates will change as a result of fully meeting the Category II requirements.

Possible Resolution To meet category II requirements, the specific surveillance procedures that are credited in section 7.3.4 should be reviewed to confirm that the components for which demands or run time are being estimated are being properly exercised within each procedure. This review should be documented in the data notebook.

Response

Four procedures were used to derive success data. In order for these procedures to be valid tests, each was reviewed to ensure that the function of the surveillance or test procedure actually tested all portions of the component within the component boundary.

For ACPFR surveillance instruction 0-SI-SXV-032-200.A (Reference 25) and 0-SI-SXV-032-200.B (Reference 26) were used to estimate the number of hours in operation in the data window. Each of these procedures requires the compressor to be in full operation for an extended period of time to assure adequate cooling to the compressor. Therefore these procedures are acceptable for use in calculating the number of run hours for ACPFR.

For AOAFC the surveillance instructions 1-SI-SXV-000-201.0 (Reference 27) and 2-SI-SXV-032-201.0 (Reference 28) were reviewed. These instructions require the valves to close within their acceptance criteria and require the complete valve to operate. Therefore these procedures are acceptable for use in calculating the number of demands for AOAFC.

F&O Number F&O Details 4-13 The PRA model currently uses flag events indicate which standby component is running. It was noted that these flags could be set to zero or one (TRUE or FALSE during quantification) as in the ERCW system, or could be set to an appropriate split Page E3-69 of 87

fraction to reflect the percentage of time each component was running as was done in the CCS system. (The base PRA model solution appears to assume a specific set of components are running or in standby, with the exception of CCS, which assumes a 50%/50% likelihood of each train running.) The mixing of assumed configurations with probabilistic configurations should be re-examined. For CCS, the split fractions are based on assumptions based on system design, but are reviewed by the system engineers for accuracy. However, operational data is not reviewed to determine the specific split fractions for each component. Therefore, the requirements of Category I for SR DA-C8 are met.

(This F&O originated from SR DA-C8)

Associated SR(s)

IE-A6 IE-C10 DA-C8 Basis for Significance This is considered to be a suggestion, as Category I of DA-C8 is met and this should be adequate for most applications. The determination of more precise split fractions (e.g., 54% and 46% vs. an assumed 50%/50% split) would not impact the overall PRA results significantly.

Possible Resolution To meet Category II, collect operating data to determine the actual standby/running fractions for plant equipment. This data could be documented either in each system notebook or in the data notebook. Consideration should also be given to using a consistent set of assumptions concerning system alignments (i.e., either all based on an assumed configuration or all based on probabilistic estimates for each alignments).

Response

The values used in each fault tree were agreed upon by the system engineers from the site. No changes will be made to the split fraction values.

Page E3-70 of 87

F&O Number F&O Details 4-14 The Level 2 documentation as presented in MDN-000-000-2010-0206 provides documentation of each of the steps of the Level 2 analysis process. The documentation is reasonably thorough; however, various documentation recommendations have been noted in the various LE SRs that will further improve the quality of the overall documentation package.

Also, there are a number of known changes to the Level 2 notebook that was reviewed that need to be made (incorporation of Sequoyah-specific containment failure evaluation results, updated CETs, updated LOSP recovery information, expanded discussion of systematic reviews that were performed, updated results, truncation studies, and sensitivity studies, etc.) This new/revised information needs to be incorporated as soon as possible in order to ensure that the notebook is consistent with the actual Level 2 model and properly documents all of the information that is required by the Standard.

(This F&O originated from SR LE-G1)

Associated SR(s)

LE-B1 LE-B2 LE-C3 LE-C9 LE-C10 LE-D1 LE-D2 LE-G1 Basis for Significance This is considered a suggestion since it pertains to documentation enhancement.

However, it is important that the Level 2 documentation reflect the current model and results.

Possible Resolution Update the notebook to reflect the comments in the referenced SRs and to incorporate the most recent information.

Response

Sequoyah specific containment failure results have been included in the document.

The containment event trees have also been updated. The LOSP recoveries use level one recovery factors, which should be conservative as discussed in Section 10.1.

Updated results, truncation and sensitivity studies have been incorporated into the current document.

Page E3-71 of 87

F&O Number F&O Details 4-15 The accident sequences are developed to a level of detail to account for the potential contributors identified in LE-B1 and analyzed in LE-B2. The containment structural capability and phenomena challenging containment are discussed in MDN-000-000-2010-0206 Sections 5.11, 5.12, and 5.13. The description of the questions considered in the Level 2 event trees are described in Section 6 and depicted in Appendix B.

However, for SBO sequences, Assumptions 10 and 11 state that if offsite power is restored, that containment systems and injection systems will then operate successfully and may arrest core melt in-vessel and provide containment heat removal. The possibility that these mitigating systems might not function should be considered following power restoration. This would affect the logic associated with questions 10 and 16 of the SBO CET.

(This F&O originated from SR LE-C1)

Associated SR(s)

LE-C1 Basis for Significance This is considered to be a suggestion since this modeling enhancement should have only a small impact on the Level 2 results (mostly for non-LERF sequences).

Possible Resolution Enhance the CET logic to consider the potential for failure of ECCS or CHR following AC power recovery.

Response

This suggestion has not been incorporated into the current model but it will be considered for future model updates. Note that while not considering failure of these systems is a source of potential non-conservative results, the current model does not credit power recoveries beyond those considered in level 1 which is a conservative counter-balance to this effect.

Page E3-72 of 87

F&O Number F&O Details 4-16 There are several issues identified with the use of the rule-based recovery file and its associated documentation. Section 9.3 of the Level 2 Notebook says that various recovery rules were used to remove invalid combinations; these have now been replaced by a combination of fault tree logic changes and mutually-exclusive rules.

The LOSP recovery information described in section 10 is out of date. Specifically, the notebook says that the additional time between core damage and vessel failure was included in the OSP recovery terms (as compared to the Level 1 analysis);

however, the Level 1 OSP recovery factors were used in the level 2 (i.e., no credit for the additional time). Also, the recovery factor data shown in the notebook pertains to Watts Bar and is not the SQN-specific data used in this PRA.

Recovery rules are also used to apply state of knowledge correlation adjustments to the ISLOCA valve failure probabilities. Similar to the issues seen with HEP dependency analysis, cutsets containing combinations of multiple valve failures may be truncated from the solution if the nominal values are used. So, an approach similar to HEP dependency rule application must be used (i.e. solve the fault tree model with artificially high values for the ISLOCA valve failures, and then use the recovery file to either apply the SOK values as appropriate or to restore the valve failure probabilities to their proper values.)

(This F&O originated from SR LE-E4)

Associated SR(s)

LE-E4 Basis for Significance This is considered to be a suggestion since the items pertain either to documentation updates or to relatively minor technical corrections. The proper treatment of the SOK adjustments may result in a small increase in ISLOCA probability. The consideration of the additional time available to restore OSP after core damage may result in a small reduction in LERF.

Possible Resolution Address the issues identified with the recovery rules.

Response

The current level 2 analysis does not use level 2 specific recovery factors and the documentation has been updated to reflect this. The difference in results produced by increasing the default failure rates in the ISLOCA tree is deemed to be very small.

ISLOCA contributes a small amount to the results for both level 1 and level 2, and utilizing a seed value approach similar to what was used for HEPS would not increase ISLOCA contribution significantly.

F&O Number F&O Details 4-17 HRA analysis was performed using the EPRI HRA Calculator for feasible operator actions following the onset of core damage. The specific operator actions credited Page E3-73 of 87

and not credited are described in MDN-000-000-2010-0206 Section 7. The details of the analysis of each HFE is contained in the HRA Notebook (MDN-000-000-2010-0204).

The SQN level 2 analysis used the Watts bar PRA as a starting point. At Watts Bar, the SG PORVs will fail upon loss of power and are not credited after that time.

However, at SQN, operators can use reach rods to operate the SG PORVs in SBO conditions. The Level 2 analysis should be updated to consider use of the SG PORVs, particularly since SBO is a key LERF contributor. Other conservative assumptions in the model (as documented in Section 5 of the Level 2 notebook) should also be re-evaluated.

(This F&O originated from SR LE-C2)

Associated SR(s)

LE-C2 Basis for Significance This is a suggestion since it could be used to improve the PRA model through the reduction of conservatism Possible Resolution Re-evaluate the manual operation of SG-PORVs in a SBO, as well as other conservative assumptions in the Level 2 analysis.

Response

The manual operation of SG-PORVs in a SBO is covered in the level 1 analysis which is incorporated in the level 2 analysis.

Page E3-74 of 87

F&O Number F&O Details 5-1 Little description of the screening values is provided in the main text.

For example, two primary screening values are used for post-initiator HEPs, 1.0E-8 and 1.0. Even though Appendix E provides the basis for the screening value, the main HRA notebook should have a description on the approach taken to assign the screening values.

In addition, Table 10-2 of the HRA notebook shows that some events in the table are set to a value of 1.0E-8. An asterisk follows this number. However, the meaning of the asterisk is not provided.

(This F&O originated from SR HR-I2)

Associated SR(s)

HR-I2 Basis for Significance This is a suggestion because these are enhancements to the documentation.

Possible Resolution

1) Document the basis of the assigned screening values and the process used to determine the appropriate values in the text of the HRA notebook.

2) Provide an explanation of the meaning of the asterisk for HEPs set to a value of 1.0E-8 in Table 10-2.

Response

1) The screening values are appropriately documented in appendix E and no value would be added by moving this section to the main text of the document.

2) Table 10-2 was revised and no HEPs are set to 1.0E-8 or have an asterisk.

Page E3-75 of 87

F&O Number F&O Details 5-4 The basis of very short manipulation time (Tm) and median response time (T1/2) values for time critical HEPs such as HAOS1 and HASE2 needs to be more completely documented. The current documentation using the Appendix C operator interview notes as a basis is not considered sufficient.

(This F&O originated from SR HR-G5)

Associated SR(s)

HR-G5 Basis for Significance This is a suggestion because this is primarily an issue with the completeness of the documentation.

Possible Resolution Perform more time evaluation, such as simulator observations, for these kinds of time critical HFEs.

Response

Various Tm and T1/2 timing values have been refined and further documentation has been added to the HRA Calculator and the calculator reports in Appendix B.

Simulator observations will be performed as they can be scheduled in the future, to further refine the HRA timing.

Page E3-76 of 87

F&O Number F&O Details 5-7 Even though MDN-000-000-2010-0201 identifies the key safety functions in general (see Section 6.2), the manner in which each key safety function is satisfied in the individual event trees is not consistently documented. For example, the discussion of the LLOCA accident sequences contains a summary of the manner in which each key safety function is modeled. However, a similar discussion is not included in the discussion of the General Transient accident sequences.

(This F&O originated from SR AS-A2)

Associated SR(s)

AS-A2 Basis for Significance The information is available in the Success Criteria Notebook, but there should be consistent discussion of the key safety functions in the various sections of the Accident Sequence Notebook.

Possible Resolution Each accident sequence should describe the key safety functions, especially for grouped initiating events like General Transients.

Response

The suggestion will be incorporated during the next revision of the notebook. For each event tree key safety functions will be explicitly detailed.

Page E3-77 of 87

F&O Number F&O Details 5-8 MDN-000-000-2010-0201 delineates all accident sequences. However the descriptions do not have clear information regarding the top events.

(This F&O originated from SR AS-C2)

Associated SR(s)

AS-C2 Basis for Significance The modeling appears to be correct, but the documentation could be improved.

Possible Resolution The event sequence description should includes a combination of failed and succeed top events. For example, the LLOCA-002 sequence description should have the event sequence using top events, LLOCA*/ACC*/LPI3*/LPR3*LPH.

Response

The accident sequence top events are described in the Accident sequence notebook, all top events have their success criteria discussed under the appropriate event tree.

Page E3-78 of 87

F&O Number F&O Details 5-11 The graphical event trees and accident sequence descriptions delineate the transfers between event trees (e.g., transfers from GTRAN to SLOCA or from SLOCA to ATWS).

However, the transferred sequence and boundary conditions are not discussed in the entry conditions for the event tree the sequence transfers to. The CAFTA single top logic method preserves the existing conditions of the transferred sequences.

Associated SR(s)

AS-A11 AS-C2 Basis for Significance The receiving event trees do not have any description about the transferred sequences from other event trees.

Possible Resolution The transfer sequences should be discussed in the initiating event discussion of the accident sequence the transfer is linked to, including the boundary conditions transferred.

Response

The boundary conditions are supplied by the input decks and their associated output files from MAAP. All transfer events from GTRAN to SLOCA are documented within the accident sequences.

Page E3-79 of 87

F&O Number F&O Details 5-15 Per Section 5.6 of MDN-000-000-2010-0208 and explanations provided by the TVA PRA analyst, the HRA event dependency in the cutset or sequence is properly assessed during the quantification process. However improvement of the documentation is needed to better explain the process by which this is done.

For example, a special software tool, "HRASeedOptimizer 2.0.0.0," was used to avoid a truncation stage from deleting necessary cutsets containing the HFE dependency combination during the quantification process. However, MDN-000-000-2010-0208 does not explain how this tool is used or how it ensures proper HEP values are used during the quantification to avoid truncation problems.

In addition, a two-stage multiplication method is used in the recovery rules for some high order HFE dependency combination due to a limitation of QRecover. However, MDN-000-000-2010-0208 does not explain the Qrecover limitation or how the applied process compensates for the limitation.

(This F&O originated from SR QU-C2)

Associated SR(s)

QU-C2 Basis for Significance The process is technically acceptable, but the documentation of the process used is incomplete.

Possible Resolution Revise the documentation to:

1) Describe the use of the HRASeedOptimizer and the manner in which it applies the selected HEP seed values.

2) Describe the Qrecover limitation requiring the use of the two-stage multiplication method and how the applied process compensates for the limitation.

Response

Added description in the Quantification notebook to address resolution in section 4.0.

Page E3-80 of 87

F&O Number F&O Details 6-1 Assessed as meeting category I. Site response to peer review team questions regarding the performance of a precursor review indicated that a review was performed, but not documented. Documentation of the review would allow for meeting capability category II.

(This F&O originated from SR IE-A9)

Associated SR(s)

IE-A9 Basis for Significance This is a suggestion for moving the categorization of this SR to Capability Category II.

Possible Resolution Document an initiating event precursor review to meet category II.

Response

Added discussion to initiating event notebook to describe how precursor events were searched for and how events for intake blockage and loss of feedwater were considered.

Page E3-81 of 87

F&O Number F&O Details 6-8 Overall, the documentation was well organized, and presented in a manner that facilitated review. There were some items that were noted that could enhance the documentation, however.

1) IE analysis, p. 17: Unit 1 results are considered applicable to Unit 2 due to similarity.

2) IE analysis, p. 17: Table A-2 of the Initiating Events Analysis refers the reader to the Accident Sequence Analysis for discussion of Supporting Requirements (SRs)

IE-B2 through IE-B5. However, Appendix A of the Accident Sequence analysis lists AS SRs but not the IE SRs.

3) Uncertainty Analysis, section 6.4: Documentation describing how initiating event frequency 5th and 95th percentile values were calculated could not be identified.

4) IE analysis, Table 5-2: The table identifies NUREG/CR-1829 as the source for initiator %LLOCACL. However, this should actually reference NUREG-1829.

(This F&O originated from SR IE-D1)

Associated SR(s)

IE-D1 Basis for Significance Items noted are enhancements to the existing documentation.

Possible Resolution Suggested resolutions to items as numbered in the description:

1) The statement on similarity could be expanded slightly to describe how the units are similar, e.g., they are similar in design, configuration and operation.

2) Provide a map to refer the reader to the documentation in which IE-B2 through IE-B5 are addressed.

3) TVA personnel indicate that Crystal Ball was used to calculate the 5th and 95th percentile values. To improve traceability of the analysis, provide a reference to this supporting analysis.

4) Correct typo.

Response

Added discussion to IE notebook to address items 1, 2, and 4.

Item 3 should be addressed in uncertainty analysis notebook.

F&O Number F&O Details 6-9 No requirements to compare results from similar plants for Capability Category I.

While a listing of CDF values from other 4 loop Westinghouse plants is provided, a Page E3-82 of 87

more thorough comparison of the results with an explanation of the causes for significant differences is required to meet Category II/III.

(This F&O originated from SR QU-D4)

Associated SR(s)

QU-D4 Basis for Significance Current documentation meets Capability Category I.

Possible Resolution Provide a more thorough comparison of PRA results beyond CDF (e.g. major sequence and initiating event contributors).

Response

A table showing the initiator distribution for similar plants has been added to the Quantification notebook.

Page E3-83 of 87

F&O Number F&O Details 6-11 Documentation of the model integration process in MDN-000-000-2010-0208 is adequate overall and contains sufficient detail to understand the process. The documentation of quantification results is also satisfactory overall. The following documentation issues were noted:

1) MDN-000-000-2010-0208, section 5.7 - The recovery events included in the importance tables do not include event descriptions.

2) MDN-000-000-2010-0208, section 5.1 references appendix F for a detailed review of the top ten cutsets - this documentation does not exist in appendix F.

3) MDN-000-000-2010-0208 Section 4.1 states 'Limitations that are known to have an impact on model development or quantification are addressed in Section 3.0 which discusses the general methodology used to develop the SQN PRA Model.' No such discussion could be found in the referenced section.

4) MDN-000-000-2010-0208 Section 4.2 indicates that the sequence successes are contained in the PRAQuant file when they are not actually contained in the PRAQuant file.

(This F&O originated from SR QU-F2)

Associated SR(s)

QU-A1 QU-B1 QU-B6 QU-F2 Basis for Significance The issues noted are minor issues that do not detract from the technical quality of the results.

Possible Resolution

1) As was done for other events in the section 5.7 importance tables, include event descriptions for the recovery events.

2) Provide the review or remove the reference.

3) Provide discussion in section 3 or delete reference.

4) Remove the referenced statement.

Response

References have been updated and all issues are resolved in the Quantification notebook.

Best Practice F&Os F&O Number F&O Details 4-5 A thorough evaluation of the potential for human-induced floods is documented in Page E3-84 of 87

section 8.6 of the flooding notebook(MDN-000-000-2010-0203). Two types of floods are considered: 'maintenance-induced floods' which considered the potential for an isolation valve to fail during a maintenance event and 'human-induced events' in which isolation is incorrectly performed, a tank is overfilled, etc. The calculated frequencies for these events are then added to the random failure frequencies for the appropriate flood initiators.

(This F&O originated from SR IFEV-A7)

Associated SR(s)

IFEV-A7 Basis for Significance This is considered a best practice, as the evaluation of human-induced events was quite comprehensive and considered different types of flooding mechanisms.

Possible Resolution N/A

Response

No response required.

Page E3-85 of 87

F&O Number F&O Details 5-10 SQN PRA delineates not only accident sequences that result in core damage (MDN-000-000-2010-0201), but also describes the success sequences (MDN-000-000-2010-0207).

(This F&O originated from SR AS-A7)

Associated SR(s)

AS-A7 Basis for Significance This is considered a best practice, as the descriptions of the accident sequences were quite comprehensive and considered both core damage and success paths.

Possible Resolution N/A

Response

No response required.

Page E3-86 of 87

F&O Number F&O Details 5-18 In addition to the general description to meet ASME standard, MDN-000-000-2010-0208 provides good information facilitating further update the model. Examples include:

- Table 7.0-1: Summary of Modeling Changes,

- Comments in recovery rule file, and

- Appendix H: steps for creating a merged model (This F&O originated from SR QU-F2)

Associated SR(s)

QU-F2 Basis for Significance This is considered a best practice, as the documentation of the changes made to the fault tree model to address system reviewers comments, modeling logic corrections, and changes based upon cutset reviews is quite comprehensive. In addition, the documentation of the recovery rules, and model integration process is quite thorough and will facilitate the model maintenance and update process.

Possible Resolution N/A

Response

No response required.

Page E3-87 of 87