ML18338A005

Regulatory Audit in Support of the License Amendment Request to Implement Risk-Informed Technical Specifications Initiative 4B

ML18338A005
Person / Time
Site:	Farley
Issue date:	01/04/2019
From:	Shawn Williams Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Williams S, NRR/DORL/LPL2-1, 415-1009
References
EPID L-2018-LLA-0210
Download: ML18338A005 (32)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 4, 2019 Ms.Cheryl A.Gayheart Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - REGULATORY AUDIT IN SUPPORT OF THE LICENSE AMENDMENT REQUEST TO IMPLEMENT RISK-INFORMED TECHNICAL SPECIFICATIONS INITIATIVE 48 (EPID L-2018-LLA-0210)

Dear Ms. Gayheart:

By letter dated July 27, 2018 (Agencywide Documents Access and Management System)

Accession No. ML18208A619), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request to implement changes to their technical specifications program that establishes a risk-informed approach for voluntary extension of completion times for Limiting Conditions of Operations based on the Nuclear Energy Institute 06-09 methodology, "Risk-Informed Technical Specification Initiative 4b Risk-Managed Technical Specification (RMTS) Guidelines," Revision 0-A, for Joseph M. Farley Nuclear Plant, Units 1 and 2.

To support its safety evaluation, the U.S. Nuclear Regulatory Commission staff will conduct an audit at the licensee's headquarters in Birmingham, AL, February 5-7, 2019. Enclosure 1 to this letter provides an audit plan in support of this audit. Enclosure 2 provides audit questions and discussion topics.

Within 45 days from the completion of the audit, the NRC staff will issue an audit summary and any formal Requests for Additional Information.

C. Gayheart If you have any questions, please contact me at 301-415-1009 or via e-mail at Shawn.Williams@nrc.gov.

Sincerely,

~r" (,) Jy)~r,r.f?__ _

Shawn A. Williams, Senior Project Manager Plant Licensing Branch II 1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

(1) Regulatory Audit Plan (2) Audit Questions cc: Listserv

REGULATORY AUDIT PLAN FOR FEBRUARY 5-7, 2019 LICENSE AMENDMENT REQUEST TO ADOPT NEI 06-09, REVISION 0-A RISK-INFORMED COMPLETION TIMES SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 BACKGROUND

By letter dated July 27, 2018 (Reference 1), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to implement changes to their technical specifications (TSs) program that establishes a risk-informed approach for voluntary extension of completion times for Limiting Conditions of Operations based on the Nuclear Energy Institute 06-09 methodology, "Risk-Informed Technical Specification Initiative 4b Risk-Managed Technical Specification (RMTS) Guidelines," Revision 0-A, for Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP).

The Nuclear Regulatory Commission (NRC) staff's review of the LAR has commenced in accordance with the Office of Nuclear Reactor Regulation's (NRR) Office Instruction LIC-101, "License Amendment Review Procedures" (Reference 2). The NRC staff has determined that a regulatory audit of the LAR should be conducted in accordance with the NRR Office Instruction LIC-111, "Regulatory Audits" (Reference 3) for the staff to gain a better understanding of the licensee's proposed risk informed completion time (RICT) program.

A regulatory audit is a planned, license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. A regulatory audit is conducted with the intent to gain understanding, to verify information, and/or to identify information that will require docketing to support the basis of a licensing or regulatory decision.

Performing a regulatory audit of the licensee's information is expected to assist the NRC staff in efficiently conducting its review and gain insights on the licensee's processes or procedures.

Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. However, the NRC staff may review supporting information retained as records and discussed in 10 CFR 50.71 "Maintenance of records, making of reports" and/or 10 CFR 54.37 "Additional records and recordkeeping requirements," which although not required to be submitted as part of the licensing action, would help the staff better understand the licensee's submitted information.

The NRC staff will conduct the audit at the licensee's headquarters in Birmingham AL, February 5-7, 2019.

The objectives of this audit are to:

• Gain a better understanding of the detailed calculations, analyses and bases underlying the LAR and confirm the staff's understanding of the LAR; Enclosure 1

• Review the approach for developing and implementing nuclear power station Risk-Managed TSs programs;

• Identify further information that is necessary for the licensee to submit in order for staff to reach a licensing or regulatory decision; discuss potential formal requests for additional information (RAls);

• Acquire understanding of the extent that the licensee's proposed amendments to modify TS requirements for RICTs are in accordance with Nuclear Energy Institute (NEI) 06-09 Revision 0-A (References 4 and 8);

• Acquire understanding whether the proposed configurations introduce any adverse effects on the ability or capacity of plant equipment to perform its design-basis function(s) when the plant is operated in the proposed TS allowable configuration; and,

• Acquire understanding of the Probabilistic Risk Assessment (PRA) quality for use in the application.

2.0 REGULATORY AUDIT BASIS The basis of this audit is the licensee's LAR for FNP, Units 1 and 2, and the Standard Review Plan (SRP) Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (Reference 5). References 5 through 8 provide additional information that will be used to support the audit.

3.0 REGULATORY AUDIT SCOPE OR METHOD The staff will review the licensee's RICT program and its application to the TSs Completion Times as proposed in the LAR. The NRC staff will review of the internal events and fire PRAs.

The team will review the risk informed approach for establishing the extended completion times.

The team will review the PRA methods used to determine the risk impact from which the revised completion times are obtained, including the internal events and fire PRAs, and the quantification of risk from significant external events through PRA or bounding methods.

The team will discuss audit questions (Enclosure 2) and identify the need for any additional information or clarification.

4.0 INFORMATION AND OTHER MATERIAL NECESSARY FOR THE AUDIT The NRC audit team will require access to personnel knowledgeable in all aspects of the licensee's LAR and the relevant PRA models (i.e., fire (F), internal events (IE), internal flood (IF)) documentation, and procedures used to support the LAR. Please have the following available to the audit team.

Documents

• Available RICT program procedures (i.e. Risk Management Action (RMA) procedure, PRA Functionality Determination Procedure, Recording Limiting Condition for Operation (LCO) Procedure, etc.);

• All PRA models (i.e., F, IE, IF, etc.) and PRA documentation should be available electronically with licensee support;

• All PRA peer review reports, self-assessments of the PRA models, and Facts and Observations (F&O) closure reports;

• Documentation of changes to the PRA models with justification of upgrades/updates;

• Plant and PRA configuration control procedures; and,

• Analyses supporting PRA success criteria, which differ from design-basis criteria.

Presentations

• PRA functional definition, development, and use, with examples;

• Configuration Risk Management Program (CRMP) demonstration

• Walkthrough sample RICT calculations;

• RMA determination with examples; and,

• Electrical systems modeling in the PRA.

Discussions

• Discuss the LAR and RICT program;

• Discuss the PRA technical adequacy and sources of uncertainty;

• Discuss how external events are considered for the RICT;

• Discuss how RMAs are determined and implemented;

• Discuss reviews and acceptance testing of the CRMP model;

• Discuss how the CRMP is maintained consistent with the baseline PRA model;

• Discuss PRA modeling for select LCOs, as identified in audit questions;

• Discuss identification of loss of function LCO conditions, as identified in audit questions;

• Discuss plant design and PRA modeling of electrical systems (offsite power, Emergency Diesel Generators (EDGs), shared or common equipment loads/systems; modeling of loads that are prioritized based on events or preference); and,

• Discuss total core damage frequency (CDF) and large early release frequency (LERF) and how cumulative risk will be evaluated and tracked.

5.0 TEAM AND REVIEW ASSIGNMENTS The on-site audit will be conducted by the following NRC staff and contractors from the Pacific Northwest National Laboratory (PNNL). Other NRC staff may participate via conference calls.

• Shawn Williams, Project Manager, Office of Nuclear Reactor Regulation (NRR), Division of Reactor Licensing (DORL)

• Anders Gilbertson, NRR, Division of Risk Assessment (ORA), Probabilistic Risk Assessment Licensing Branch A (APLA)

• Brandon Hartle, NRR, ORA, APLA

• Mihaela Biro, NRR, ORA, APLA

• Gurcharan Matharu, NRR, Division of Engineering (DE), Electrical Engineering Operating Reactors Branch (EEOB)

• Margaret Chernoff, NRR, Division of Safety Systems (DSS), Technical Specifications Branch {STSB)

• Garill Coles, PNNL (Contractor)

The NRC technical leads will be Brandon Hartle (PRA) and Anders Gilbertson (PRA). The audit team leader will conduct daily briefings on the status of the review and coordinate audit activities while on site. The table below shows the planned audit team composition and their assigned areas for review during the audit.

Regulatory Audit Plan Review Areas and Assignments Audit Plan Review Areas Lead Suooort 1 Engineering evaluations, previous approval Team Team 2 Risk Assessment APLA PNNL 3 External Hazards Risk Contribution APLA/B PNNL 4 PRA Technical Adequacy APLA PNNL 5 Uncertainty/Assumptions APLA PNNL 6 Documentation, Configuration Control, Quality APLA PNNL 7 Risk Management Actions APLA PNNL 8 Monitoring program APLA PNNL 9 PRA Functionality APLA PNNL 10 Electrical Systems design and modeling APLA/ EEOB PNNL 11 Plant-specific Technical Specifications STSB PNNL 6.0 AUDIT MILESTONES AND SCHEDULE Audit Milestones and Schedule Activity Timeframe Comments 1-3 weeks Teleconference from NRC headquarters to provide Clarification Call prior to the clarification of audit questions Audit Brief team introduction and discussion of the scope Onsite Audit Kickoff 2/5/2019 of the audit. The licensee should introduce team Meeting members and give logistics for the week.

Meet with licensee to provide a summary of any End of Day Summary 2/5/2019-significant findings and requests for additional Briefings 2/7/2019 assistance.

Provide Rooms for Facilitate discussions between site and staff 2/5/2019-Focused Topic technical areas. Provide one or two breakout 2/7/2019 Discussions areas, if possible, for smaller discussions.

Onsite Audit Exit NRC staff will hold a brief exit meeting with the 2/7/2019 Meeting licensee's staff to conclude audit activities.

Audit Summary (see < 45 days after Document the audit and any requests for additional Section 9.0) exit information.

7.0 LOGISTICS This regulatory audit will begin February 5, 2019, and will last approximately 3 days. A conference call with the licensee will be scheduled 1 to 2 weeks prior to the audit to discuss the details of the audit plan. The dates in the milestone chart are subject to change based on mutual agreement between the licensee and the NRC. An entrance meeting for this audit will be held on the first day at 8:00 a.m. and an exit meeting will be held the final day no later than 5:00 p.m. (or at another time mutually agreed to by the licensee and the NRC). The NRC audit team leader will provide daily progress briefings to licensee personnel on the first and second day of the audit.

The audit will take place at SNC headquarters located at 3535 Colonnade Parkway, Birmingham, AL. Visitor access will be requested for the entire audit team. It is recommended any security paperwork and processing be handled prior to the first day of the audit, if possible.

The audit will start at 8:00 a.m. (Central Standard Time (CST)) on Tuesday, February 5, and conclude on Thursday, February 7, at approximately 5:00 p.m. (CST).

8.0 SPECIAL REQUESTS If available, the regulatory audit team requests the following to support the regulatory audit:

• One main conference room (that can hold approximately 12 staff) with one additional private area for conference calling capability and access to visitor wifi.

• Access to licensee personnel knowledgeable in the plant design, operation and the plant PRAs. In addition, FNP staff who participated in the internal events and fire F&O closure process and in preparing the LAR should be available for discussion.

• Access to internet-based portal provided by the licensee to access the Farley Internal Events Probabilistic Risk Assessment Peer Review F&O Closure Report; Farley Fire Probabilistic Risk Assessment Peer Review F&O Closure Report, and SNC responses to NRC Staff questions documented in the November 7, 2018, Meeting Summary (ADAMS Accession No. ML18306A313). NRC staff requests access to the portal until the LAR is complete.

9.0 DELIVERABLES Within 45 days from the completion of the audit, the NRC staff will issue an audit summary and any formal Requests for Additional Information. The audit summary will use the guidance of NRR Office Instruction LIC-111 for content.

10.0 REFERENCES

1. Letter from Cheryl A. Gayheart, Director, Regulatory Affairs, Southern Nuclear Operating Company, to U.S. Nuclear Regulatory Commission, "Joseph M. Farley Nuclear Plant Units 1 and 2, License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0-A, 'Risk-Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines'," dated July 27, 2018 (ADAMS Accession No. ML18208A619).

2. U.S. Nuclear Regulatory Commission (NRC), NRR Office Instruction, LIC-101, Revision 4, "License Amendment Review Procedures," May 22, 2012 (ADAMS Accession No. ML113200053).

3. U.S. NRC, NRR Office Instruction, LIC-111, "Regulatory Audits," December 29, 2008 (ADAMS Accession No. ML082900195.

4. Nuclear Energy Institute, NEI 06-09, "Risk-Informed Technical Specification Initiative 4b Risk-Managed Technical Specification Guidelines," Revision 0, November 2006 (ADAMS Accession No. ML063390639).

5. U.S. NRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (ADAMS Accession No. ML071700658).

6. U.S. NRC, Regulatory Guide 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated January 2018 (ADAMS Accession No. ML17317A256).

7. U.S. NRC, Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009 (ADAMS Accession No. ML090410014).

8. U.S. NRC, "Final Safety Evaluation for Nuclear Energy Institute Topical Report 06-09,

'Risk-Informed Technical Specification Initiative 4b Risk-Managed Technical Specification Guidelines'," Revision 0, May 2007 (ADAMS Accession No. ML071200238).

AUDIT QUESTIONS LICENSE AMENDMENT REQUEST TO ADOPT NEI 06-09, REVISION 0-A RISK-INFORMED COMPLETION TIMES SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364 By letter dated July 27, 2018 (Reference 1) Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to modify the Joseph M. Farley Nuclear Plant (Farley) Technical Specification (TS) requirements to permit the use of Risk-Informed Completion Times (RICT) in accordance with Nuclear Energy Institute (NEI) 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference 2).

The U.S. Nuclear Regulatory Commission (NRC) staff has determined that a regulatory audit of the LAR should be conducted in accordance with the Nuclear Reactor Regulation (NRR) Office Instruction LIC-111, "Regulatory Audits" (Reference 3) for the staff to gain a better understanding of the licensee's proposed RICT program.

The following is a list of audit questions for discussion to aid in the development of any formal Requests for Additional Information (RAI). If the NRC staff determines that the information in an audit question is needed to reach a licensing or regulatory decision, the NRC will issue a formal RAI within 45 days of the audit.

Question No. 1 - Internal Events, Internal Flood, and Internal Fire Probabilistic Risk Assessment (PRA) Facts and Observations (F&O)

Regulatory Guide (RG) 1.200, Revision 2 (Reference 4) provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process using the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA 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 5), as one acceptable approach for determining the technical acceptability of the PRA. The primary result of a peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the NEI guidance documents NEI 05-04, NEI 07-12, and NEI 12-13, titled "NEI 05-04/07-12/12-06 Appendix X: Close-out of Facts and Observations" (Reference 6 and 7),

which was accepted by the NRC in a letter dated May 3, 2017 (Reference 8).

Enclosure 2

An F&O closure review meeting was conducted in October 2018 at the SNC offices, which the NRC staff observed, to close out the F&Os associated with the Farley internal events PRA (IEPRA) and internal flood PRA (IFPRA). Additionally, an F&O closure review was completed in September 2018 for the Farley internal fire PRA (FPRA) model.

Questions/Discussions:

a. Summarize the results of the F&O closure reviews and update the dispositions for any finding-level F&Os not closed by the F&O closure review by explaining how the F&Os are resolved for this application.

b. Confirm that all finding-level F&Os, including findings against PRA supporting requirements that were met at Capability Category II (CC II), were provided to the independent assessment (IA) team for the F&O closure reviews.

c. Appendix X guidance states in part, "[a]dditionally, the team will review the SR

[Supporting Requirement] to ensure that the aspects of the underlying SR that were previously not met, or met at CCI, are now met, or met at CCII.

Confirm that closure of the F&Os was assessed to ensure that the capabilities of the PRA elements, or portions of the PRA within the elements, associated with the closed F&Os now meet all ASME/ANS RA-Sa-2009 SRs at CC-II.

d. Describe whether resolution to any F&O was found to meet the definition of a PRA upgrade consistent with the ASME/ANS RA-SA-2009 PRA standard.

e. For any upgrade, confirm that a focused-scope peer review was conducted and provide all F&Os that resulted from the focused scope peer review along with their associated dispositions for the application.

Question No. 2 - Incorporation of IEPRA F&O Resolutions into the FPRA RG 1.200, Revision 2, describes a peer review process using an associated ASME/ANS PRA standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical acceptability of the PRA. Supporting requirement PRM-82 of ASME/ANS PRA standard states that a verification should be performed to confirm that deficiencies in the IEPRA model do not adversely affect development of the FPRA model.

Prior to the IEPRA and IFPRA F&O closure review, all open finding-level FPRA F&Os were closed in a FPRA F&O closure review that was initiated in April 2018 and finalized in September 2018. The IEPRA model forms the basis for the FPRA plant response model, and therefore, Findings in the IEPRA can also impact the adequacy of the FPRA model.

Questions/Discussions:

a. Identify any IEPRA modeling updates that were not incorporated into the FPRA, and provide justification regarding why these updates do not apply to the FPRA or do not have an impact on the application.

b. As an alternative to part (a) above, propose a mechanism that ensures all applicable

IEPRA modeling updates that were performed to resolve F&Os for the F&O closure review are incorporated into the FPRA prior to implementation of the RICT program.

Question No. 3 - Credit for NFPA 805 Plant Modifications Section C.2.3.4 of RG 1.174, Revision 3 (Reference 9), states that the PRA results used to support an application should be derived from a PRA model that represents the as-built and as-operated plant to the extent needed for the application. Consistent with this regulatory position, the PRA should realistically reflect the risk associated with the plant at the time a RICT is credited. Section 4 of NRC's safety evaluation (SE) (Reference 10) of NEI 06-09 states, in part, that, "The LAR will provide a discussion of the licensee's programs and procedures which assure the PRA models which support the RMTS are maintained consistent with the as-built, as-operated plant."

According to the amendment related to the NFPA 805 Supplement, as approved by the NRC on November 1, 2017 (Reference 11 ), the licensee is required to implement all the plant modifications, before the conclusion of the Farley Unit 1 spring 2018 Refueling Outage (1R28).

Questions/Discussions:

a. Confirm that all plant modifications and implementation items, such as updating the fire response procedures, credited in the FPRA associated with transition to the NFPA 805 program have been completed.

b. If all credited plant modifications and implementation items associated with transition to the NFPA 805 program are not incorporated into the FPRA model, then propose a mechanism that ensures that the FPRA model used in the RICT program does not credit plant modifications or implementation items that have not been completed.

c. If, as result of the NFPA 805 Implementation Items 30 and 32 (References 12 and 13),

the FPRA has been updated to reflect the as-built, as-operated plant associated with the final completed NFPA 805 modifications and procedures, then propose a mechanism that ensures that the FPRA is updated to reflect the as-built, as-operated plant prior to implementation of the RICT program.

Question No. 4 - Potential Credit for Flexible Coping Strategies (FLEX} Equipment or Actions The NRC memorandum dated May 30, 2017, "Assessment of the Nuclear Energy Institute 16-06, 'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis" (Reference 14), provides the NRC's staff assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2. Though implementation of FLEX procedures is cited in the LAR as possible Risk Management Actions (RMAs), the LAR and other docketed information do not indicate that SNC has credited FLEX equipment or actions into their internal events or FPRA models.

Questions/Discussions:

a. State whether FLEX equipment and strategies have been credited in the PRA. If not incorporated or their inclusion is not expected to impact the PRA results used in the categorization process, no additional response is requested.

b. If the equipment or strategies have been credited, and their inclusion is expected to impact the PRA results used in the categorization process, please provide the following information separately for IEPRA, external hazards PRA, and external hazards screening as appropriate:

i. A discussion detailing the extent of incorporation, i.e. summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to support this application.

ii. A discussion detailing the methodology used to assess the failure probabilities of any modeled equipment credited in the licensee's mitigating strategies (i.e., FLEX). The discussion should include a justification explaining the rational for parameter values, and whether the uncertainties associated with the parameter values are considered in accordance with ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Revision 2.

iii. A discussion detailing the methodology used to assess operator actions related to FLEX equipment and the licensee personnel that perform these actions. The discussion should include:

1. A summary of how the licensee evaluated the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-U) of supporting requirement HR-G3 of ASME/ANS RA-Sa-2009.

2. Whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that renders the equipment unavailable during an event, and if the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of ASME/ANS RA-Sa-2009.

3. If the licensee's procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, provide a discussion detailing the technical bases for probability of failure to initiate mitigating strategies.

c. ASME/ANS RA-Sa-2009 defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences.

Section 1-5 in Part 1 of ASME/ANS RA-Sa-2009 states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard.

Questions/Discussions:

i. Provide an evaluation of the model changes associated with incorporating mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, (3) change in

capability that impacts the significant accident sequences or the significant accident progression sequences, OR ii. Propose a mechanism to ensure that a focused-scope peer review is performed on the model changes associated with incorporating mitigating strategies, and associated F&Os are resolved to Capability Category II prior to implementation of the RICT program.

Question No. 5 - Modeling of the Reactor Coolant Pump (RCP) Shutdown Seals LAR, Enclosure 7, states that the plant has been modified to install the RCP shutdown seals and that the shutdown seal is modeled consistent with WCAP-17100-NP, "PRA Model for the Westinghouse Shut Down Seal," Revision 1 (Reference 15), which applies to Generation I and II. In the licensee's October 12, 2017, response to NRC Request for Additional Information (RAls) (Reference 16), the licensee states that the PRA includes credit for the Westinghouse Generation Ill RCP seals.

The PRA model for the Generation Ill Seals was approved by the NRC in the final safety evaluation of Topical Report (TR) PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal," dated the August 23, 2017 (Reference 17).

Consistent with the guidance in RG 1.174, Revision 3, that the PRA scope, level of detail and technical acceptability be based on the as-built and as-operated and maintained plant, and reflect operating experience at the plant.

Questions/Discussions:

a. Clarify what kind of seals are installed in each RCP in FNP Unit 1 and 2 and whether the current IEPRA and FPRA models include credit for the Westinghouse Generation Ill

("SHIELD") RCP seals.

b. If the IEPRA or the FPRA models include credit for the Westinghouse Generation Ill RCP seals, address the following:

i. Confirm that the limitations and conditions in the NRC safety evaluation for PWROG-14001-P, Revision 1, are met.

ii. If exceptions to the limitations and conditions exist, identify all the exceptions and justify their impact on the application.

iii. Clarify whether the Generation Ill Westinghouse RCP seal model has been peer-reviewed as part of the IEPRA and FPRA peer-reviews.

iv. If this RCP seal model has not been peer reviewed, justify why the addition of this model is not considered a PRA upgrade requiring a focused-scope peer review.

v. If the addition of RCP seal model qualifies as a PRA upgrade, provide the results from the focused-scope peer review including the associated F&Os and their resolutions.

c. If the PRA models do not include credit for the failure of any RCP seal, provide justification for not modeling the seals for the application.

Question No. 6 - Supplanted and Updated Fire PRA Guidance Since the safety evaluation (SE) was issued to Farley for implementation of its National Fire Protection Association (NFPA) 805 program, NRC has issued updated guidance for aspects of FPRA that supplant earlier guidance issued by NRC. Recently, as part of NRC's review of SNC's request to revise the Farley integrated leak rate test (ILRT) program, NRC staff requested information about the impact of such guidance on the ILRT application in a letter dated March 15, 2017 (Reference 18). Specifically, NRC requested information about the impact of the following NRC FPRA guidance documents.

• NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE)"

(Reference 19) regarding the updated approach to credit incipient fire detections systems;

• NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database" (Reference 20) regarding changes in fire ignition frequencies and non-suppression probabilities;

• NUREG/CR-7150, "Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE)," Volume 2, "Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure" (Reference 21) regarding possible increases in spurious operation probabilities.

In response to NRC's request, SNC provided the results of a sensitivity study in a letter dated October 12, 2017, demonstrating that the aggregate impact from applying the new guidance was an increase in total core damage frequency (CDF) of about 32 percent for Unit 1 and an increase in CDF of about 19 percent for Unit 2. Enclosure 4 of the LAR shows that fire risk is the dominant risk contribution to total risk by a significant margin (e.g., the fire CDF for Unit 1 is about a ten times greater than the internal events CDF for unit 1). Accordingly, the sensitivity study results demonstrate that the aggregate impact of using the updated NRC guidance cited above could impact the RICT calculations.

Questions/Discussions:

a. Clarify whether or not guidance from the cited guidance documents (i.e., NUREG-2180, NUREG-2169, and NUREG/CR-7150) has been incorporated into the FPRA that will be used to support the RICT program.

b. If the guidance from the guidance documents cited above (i.e., NUREG-2180, NUREG-2169, and NUREG/CR-7150) has not been incorporated into the FPRA, then propose a mechanism that ensures that the guidance from NUREG-2180, NUREG-2169, and NUREG/CR-7150 is incorporated into the FPRA prior to implementation of the RICT program.

Question No. 7 - Sources of Uncertainty - Seasonal Variations of the LAR, which identifies sources of PRA modelling uncertainty, states that no

internal events or FPRA modeling uncertainties were identified that require a sensitivity study as part of the RICT program calculations. However, NRC staff identified a source of uncertainty that appeared to have the potential to impact the application. LAR Enclosure 7, Section 2 states that the baseline PRA does not account for seasonal variations caused by external hazards even though "certain initiating events can be affected by seasonal variations." It further states:

The RICT Program will include a qualitative consideration of weather events as part of the RMA decision process when LCO 3.8.1 [AC Sources] CTs are extended to address this source of uncertainty.

Section 2.3.4 of NEI 06-09, Revision 0-A, states, in part:

If the PRA model is constructed using data points or basic events that change as a result of time of year or time of cycle (examples include moderator temperature coefficient, summer versus winter alignments for HVAC, seasonal alignments for service water), then the RICT calculation shall either 1) use the more conservative assumption at all time, or 2) be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration for the feature as modeled in the PRA. Otherwise, time-averaged data may be used in establishing the RICT.

NRC staff is unclear how the qualitative consideration of weather events will be used as part of the RMA decision process. Additionally, the NRC notes that it is possible for other initiating events (e.g., loss of Plant Service Water) to also be impacted by seasonal variations in external hazards.

Questions/Discussions:

a. Describe how the RMA process will use qualitative considerations of weather events when LCO 3.8.1 completion times are extended to address uncertainties associated with seasonal variations in external hazards. Include in this description whether the qualitative consideration will be used to adjust initiating event frequencies based on seasonal variations and how the proposed use of qualitative considerations of weather events is consistent with the guidance in NEI 06-09, Revision 0-A.

b. Explain and justify whether initiating events other than loss of offsite power included in the Configuration Risk Management Program (CRMP) modeling can be impacted by seasonal variations in external hazards.

c. If initiating events other than loss of offsite power used in the CRMP model can be impacted by seasonal variations in external hazards, then identify these initiating events and justify why seasonal variation in initiating event frequency cannot impact the RICT calculations.

d. As an alternative to part (c), identify initiating events whose event frequencies can be impacted by seasonal variations in external hazards and the LCOs that can be impacted and confirm that the RICT program will qualitatively consider those events during the RMA development process when the completion time of impacted LCOs are extended.

Question No. 8 - PRA Modeling Conservatisms That May Result in Underestimation of a RICT The NRC SE of NEI 06-09, Revision 0-A states:

When key assumptions introduce a source of uncertainty to the risk calculations (identified in accordance with the requirements of the ASME standard), TR NEI 06-09, Revision 0, requires analysis of the assumptions and accounting for their impact to the RMTS calculated RICTs. of the LAR states that no IEPRA or FPRA modeling assumptions or uncertainties were identified that require a sensitivity study as part of the RICT program calculations.

However, NRC notes that a couple of assumptions identified in LAR Enclosure 7 appear to have the potential to impact the RICT calculations for certain structures, systems and components (SSCs). Though these assumptions are described as conservative assumptions, NRC staff notes that conservatism in PRA modeling could have a non-conservative impact on the RICT calculations for certain SSCs. If a SSC is part of system not credited in the FPRA or it is supported by a system that assumed to always fail, then the risk increase due to taking that SSC out of service could be masked by the conservative modeling.

Questions/Discussions:

a. An entry in LAR Table E?.3 indicates that "some systems are not credited" in the FPRA by treating them as always failed because the associated SSC cables were not traced.

This conservative modelling can mask the risk associated with taking certain SSCs out of service and impact on the calculated RICT. It appears to the NRC staff that this uncertainty may require sensitivity studies to support the RICT calculations.

i. Identify the systems that are assumed always to be failed in the FPRA and explain whether Farley's proposed RICT program includes secs that are part of systems are not credited in the FPRA or are supported by a system that is not credited in the FPRA.

ii. If the RICT program includes SSCs that are part of system not credited in the FPRA or are supported by a system that is not credited in the FPRA, then justify that the uncredited systems have an inconsequential impact on the RICT calculations.

Alternatively, propose a mechanism to ensure that a sensitivity is performed for the RICT calculations for applicable SSCs to determine the impact of the conservative modeling on the RICT. The proposed mechanism should also ensure that any additional risk associated with the modeling is either accounted for in the RICT calculation or is compensated for using an additional RMA during the RICT.

b. LAR Table E7.1, associated with sources of modelling uncertainty for the IEPRA, states that credit for battery life was limited to "two hours based on conservative FSAR analysis." The disposition for this source of uncertainty states that this uncertainty is "unlikely to be an issue for delta risk applications" citing possible manual actions that could be taken if DC power is lost. However, it is not clear to NRC staff that this assumption, which excludes credit for proceduralized actions that would extend battery life, does not impact the application. This conservative modelling can mask the risk associated with taking certain SSCs out of service (e.g., a DC electrical power subsystem) and entry into previously unanalyzed procedures. All of which have a

potential impact on the calculated RICT. It appears to NRC staff that this uncertainty may require sensitivity studies to support the RICT calculations.

i. Justify that the exclusion of credit for actions that would extend battery life has an inconsequential impact on the application including RICTs calculated for LCOs that involve the operability of DC electrical power SSCs.

ii. If, in response to part (i) above, it cannot be justified that conservative modeling of battery life does not result in underestimation of the RICT for certain LCOs such as those associated with DC functionality, then propose a mechanism to ensure that a sensitivity is performed as part of the RICT calculations associated with those impacted LCOs to determine the impact of the uncertainty on the RICT. The proposed mechanism should also ensure that any additional risk associated with the modeling uncertainty is accounted for in the RICT calculation, or that an additional RMA is applied during the RICT.

Question No. 9 - Sources of Model Uncertainty and Parametric Uncertainty Methodology Section C of RG 1.174, Revision 3, states, in part:

In implementing risk-informed decisionmaking, LB [licensing basis] changes are expected to meet a set of key principles .... In implementing these principles, the staff expects [that]: ... Uncertainty receives appropriate consideration in the analyses and interpretation of findings .... NUREG-1855 provides acceptable guidance for the treatment of uncertainties in risk-informed decisionmaking. of the LAR states that the IEPRA uncertainty analysis was performed based on guidance from NUREG-1855 and cites Revision O in the Reference section of Enclosure 7, Section 2.0. Revision O of NUREG-1855 (2009) primarily addressed sources of model uncertainty for internal events (including internal flood). Revision O of NUREG-1855 EPRI report 1016737, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments" (Reference 22), which, among other guidance, provides a generic list of sources of model uncertainty and related assumptions for internal events. Revision 1 of NUREG-1855 (Reference 23) further clarifies the NRC staff decisionmaking process in addressing uncertainties and addresses all hazard groups (e.g., internal events, internal flood, internal fire, seismic, low-power and shutdown, Level 2). NUREG-1855, Revision 1, cites use of Electric Power Research Institute (EPRI) 1016737 and EPRI 1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty" (Reference 24), as companion documents to the NUREG-1855 that provide generic lists of sources of model uncertainty for internal events, internal flood, internal fire, and other hazard groups.

In addition, based on RG 1.174, Revision 3, and Section 6.4 of NUREG-1855, Revision 1, for a CC II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to are the

• means of the risk metric's probability distributions that result from the propagation of the uncertainties on the PRA input parameters and those model uncertainties explicitly represented in the model. In general, the point estimate CDF and large early release frequency (LERF) values obtained by quantification of the cutset probabilities using mean values for each basic event probability does not produce a true mean of the CDF/LERF.

Finally, under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the state-of-knowledge correlation (SOKC) is unimportant (i.e., the risk results are well below the acceptance guidelines).

Questions/Discussions:

a. Describe the process used to identify the LAR sources of model uncertainty and related assumptions, including generic and plant-specific sources, in the Farley baseline IEPRA, IFPRA, and FPRAs and that were evaluated for their potential impact on this application.

Include in this discussion an explanation of how the process is consistent with guidance in NUREG-1855, Revision 1, and the complementary EPRI documents 1016737 and 1026511, or other NRC-accepted methods.

b. Identify any sources of model uncertainty and related assumptions provided in NUREG-1855, Revision 1, EPRI 1016737, and EPRI 1026511 that were not evaluated in the LAR for relevance to the application.

c. For those sources of model uncertainty and related assumptions that could impact the application:

i. Provide qualitative or quantitative justification that these key uncertainties and assumptions do not cause the baseline PRA results to challenge the RG 1.174, Revision 3, risk acceptance guidelines. (NRC staff notes that given the total CDF and LERF results presented in LAR Enclosure 4, there is a small margin between the total CDFs for Farley Units 1 and the RG 1.174 risk acceptance guidelines.)

ii. Justify that these key uncertainties and assumptions have no impact on the RICT calculations. Alternatively, propose a mechanism to ensure that a sensitivity is performed for the uncertainties when a RICT calculation is performed to assess the impact of the uncertainty on the RICT. As part of the proposal, ensure that the additional risk is accounted for in the RICT calculation or that an additional RMA is be applied during the RICT.

d. Clarify whether the total CDF and LERF values presented in LAR Enclosure 4 are mean values, consistent with the guidance in NUREG-1855, Revision 1, and RG 1.174, Revision 3. Also confirm whether the total CDF and LERF values presented in LAR Enclosure 4 are derived from the results of a parametric uncertainty analysis that considers the SOKC in the IEPRA, IFPRA, and FPRA.

e. If risk values were not estimated consistent with guidance in NUREG-1855, Revision 1, and RG 1.174, Revision 3, as described above, then confirm the RG 1.174 total risk acceptance guidelines are still met when mean values are used and the SOKC is accounted for in the PRA models' results.

Question No. 10 - Total Risk Estimates Against RG 1.174 Guidelines RG 1.174, Revision 3, provides the risk acceptance guidance for total CDF ( 1x104 per year) and LERF ( 1x10-5 per year). Enclosure 4 of the LAR shows the total CDF for Unit 1 to be 9.69x10-5 per year and for Unit 2 to be 9.22x10-5 per year), thus demonstrating a small margin between the total unit risk and the RG 1.174 risk acceptance guidelines. NRC staff notes that

the response to a number of the preceding questions (i.e., 1, 2, 3, 6 and 9.e) could involve updates to the IEPRA, IFPRA, or FPRA models.

Questions/Discussions:

a. Demonstrate that after the IEPRA, IFPRA, and FPRA models are updated in response to the cited questions that the total risk for each unit is recalculated from the updated models and confirmed to still be in conformance with RG 1.174, Revision 3, risk acceptance guidance (i.e., a CDF and LERF less than 1x10-4 and 1x10-5 per year, respectively).

b. Alternatively, propose a mechanism ensuring that after the IEPRA, IFPRA, and FPRA models are updated in response to questions and prior to implementation of the RICT program, the total risk for each unit is recalculated from the updated models and confirmed to still be in conformance with risk acceptance guidance in RG 1.174, Revision 3.

Question No. 11 - Screening of External Hazards Section 2.3.1, Item 7, of NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A, states that the "impact of other external events risk shall be addressed in the RMTS program," and explains that one method to do this is by documenting prior to the RMTS program that external events that are not in modeled in the PRA are not significant contributors to configuration risk. The SE for NEI 06-09, states that "[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk."

Section 1.2.5 of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities," Revision 2, states that the contribution of many external events to CDF and LERF can be screened out "(1) if it meets the criteria in NRC's 1975 Standard Review Plan (SRP) or later revision; or (2) if it can be shown using a demonstrably conservative analysis that the mean value of the design-basis hazard used in the plant design is less than 10-5 per year and that the conditional core damage probability is less than 10 -1 , given the occurrence of the design-basis-hazard event; or (3) if it can be shown using demonstrably conservative analysis that the CDF is less than 1o-6 per year." The screening criteria listed in Section 1.2.5 of RG 1.200 are consistent with those in Section 6-2.3 of the 2009 American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA Standard (RA-Sa-2009), "Addenda to ASMEIANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications."

External Flooding of Attachment 2 to the LAR discusses the evaluation of external event challenges.

Table E3.A2.1 of the same enclosure discusses the basis for screening the external flooding hazard from consideration for this application. The basis for screening the external flooding hazard includes the results documented in the licensee's flood hazard reevaluation report (FHRR; Reference 25). In addition, the licensee states that the combined effects river flooding event is estimated to produce a maximum flood elevation that will not top the vehicle barrier system (VBS).

The licensee's basis for screening the external flooding hazard from consideration for this

application seems to rely on compliance with the Current Design Basis (CDB) with respect to mitigation of the hazard, and the use of deterministic theoretical maximum values for flood elevations.

However, the licensee's basis does not address the frequency of exposure to flood hazards (including occurrences lower than the design basis) that may impinge upon SSCs and challenge plant safety, the impact of associated effects and the temporal characteristics of the event (e.g.,

the period of site inundation), and the reliability of flood protection features and human actions.

For example, the VBS appears to be an SSC whose ability to mitigate an external flooding event, based on its design and function, is not guaranteed.

Based on the FHRR, the licensee states in Table E3.A2.1 of Enclosure 3 of Attachment 2 to the LAR that the frequency of a local intense precipitation (LIP) event capable of producing flood magnitudes reported in the FHRR is estimated to be well below 1x1 o-e per year without providing the basis for the frequency estimation. The FHRR does not provide any estimation of hazard frequency. The staff notes that Section 6.2-3 of the 2009 ASME/ANS PRA Standard as endorsed in RG 1.200, Revision 2, discusses the importance of recognizing that the demonstratively conservative estimate of a mean value is not a point estimate because the mean frequency can fall above the 95 percentile of the distribution when uncertainties are large.

Questions/Discussions:

a. Provide justification using the criteria in Section 6.2-3 of ASME/ANS RA-Sa-2009 for screening the external flooding hazard from this application. The justification should include consideration of uncertainties in the determination of demonstrably conservative mean values as discussed in Section 6.2-3 of ASME/ANS RA-Sa-2009.

b. If the external flooding hazard cannot be screened out in item (a), discuss, using quantitative or qualitative assessments, how the risk from external flooding hazards, especially the LIP and combined events river flooding, will be considered in the risk-informed completion times (RICTs) that are impacted by those*hazards. The discussion should include consideration of and, as applicable, the basis for the following factors:

• The frequency of LIP and combined events river flooding hazards,

• The impact of LIP and combined events river flooding on plant operation and structures including the ability to cope with upset conditions,

• The reliability of flood protection measures, and

• The reliability of operator actions.

c. If the external flooding hazard is screened out in item (a), discuss how it will be ensured that assumptions related to the availability and the functionality of flood protection features (e.g., VBS) that are credited for the screening remain valid during RICTs such that the external flooding hazard continues to have an insignificant impact on the configuration-specific risk.

High Wind and Tornados of Attachment 2 to the LAR discusses the evaluation of external event challenges.

Table E3.A2.1 of the same enclosure discusses the basis for screening the extreme winds and tornados (including generated missiles) from consideration for this application. The licensee's basis for screening relies on the design of SSCs and a detailed tornado missile risk analysis.

The discussion further states that "the site is currently evaluating tornado missiles in response to Regulatory Issue Summary (RIS) 15-06 (Reference 26) and that the results of that evaluation "will be reflected in the extreme winds and tornados screening evaluation."

The licensee states in Table E3.A2.1 of Enclosure 3 of Attachment 2 to the LAR that the frequency of missile damage to target groups is less than 7x10-1 per year per Unit. The staff notes that the licensee's site is located in NRC's tornado region I as shown in RG 1.76, Revision 1 (Reference 27). Additionally, the evaluation performed in response to RIS 15-06 is focused on identified non-compliances against the design basis tornado missile protection. The criteria in Section 6.2-3 of the 2009 ASME/ANS PRA Standard does not appear to be have been considered in the evaluation for high winds and tornados (i.e., impacts other than tornado missile risk) in Table E3.A2.1 of Enclosure 3 of Attachment 2 to the LAR.

d. Provide justification using the criteria in Section 6.2-3 of ASME/ANS RA-Sa-2009 for screening high wind and tornados hazard (i.e., impacts other than tornado missile risk) from this application. The justification should include consideration of uncertainties in the determination of demonstrably conservative mean values as discussed in Section 6.2-3 of ASME/ANS RA-Sa-2009.

e. If the high winds and tornados hazard cannot be screened out in item (a), discuss, using quantitative or qualitative assessments, for how the risk from high wind and tornado hazards will be considered in the RICTs that are impacted by those hazards. The discussion should include consideration of, as applicable, the basis for the following factors:

• The frequency of high winds and tornados,

• The impact of high winds and tornados on plant operation and structures including the ability to cope with upset conditions, and

• The reliability of operator actions.

f. Discuss the approach used to obtain the tornado missile damage frequency of 7x10-1 per year and the appropriateness of that risk analysis to support the screening of the tornado missile risk for the RICTs affected by tornado missile impact.

g. The statement "[r]esults of the [tornado missile protection] TMP evaluation will be reflected in the extreme winds and tornados screening evaluation" in Table E3.A2.1 of Enclosure 3 of Attachment 2 to the LAR implies future consideration of that evaluation for screening the tornado missile risk for the current application. Explain the intent of the cited statement and how results of future evaluations and staff decisions can be factored into the current application. If the TMP evaluation is expected to be used to support the screening for tornado missile hazard for the current application, discuss the approach used for that evaluation, its appropriateness to support the screening of the tornado

missile risk for the RICTs affected by tornado missile impact, and differences compared to the approach discussed in item (c).

Question No. 12 - License Condition LAR, Attachment 1, Section 4 states that to ensure changes in PRA methods are addressed wording similar to the following will be adopted in the License Condition:

The risk assessment approach and methods, shall be acceptable to the NRC, be based on the as-built, as-operated, and maintained plant, and reflect the operating experience of the plant as specified in RG 1.200. Methods to assess the risk from extending the completion times must be PRA methods accepted as part of this license amendment, or other methods approved by the NRC for generic use. If the licensee wishes to change its methods, and the change is outside the bounds of this license condition, the licensee will seek prior NRC approval via a license amendment.

LAR, Attachment 1, Section 4 also states that LAR Attachment 5 contains the marked-up and clean pages for the operating license with this particular condition included. However, was not submitted as part of the LAR dated July 27, 2018.

Question/Discussion:

Provide the changes to the operating license that with the license condition that contains the wording cited above.

Question No. 13 - LCO 3.6.2 C, Containment Air Locks The LAR Enclosure 1, Table E1 .1 regarding LCO 3.6.2 Condition C, (One or more Containment Air Locks inoperable) indicates that the SSCs are modeled consistent with the TS scope and so can be directly evaluated using the CRMP tool. It also states that the PRA success criteria are consistent with the design basis success criteria. The same table indicates that the design basis success criterion is "Post-Accident Containment Leakage Rates within limits." NRC notes that system success criteria is typically modelled in a PRA by using fault tree logic to define how many components or trains are needed for success.

Questions/Discussions:

a. Explain how the Containment Air Locks are modeled in the PRA model (CRMP model) supporting the RICT program to reflect the design basis success criteria and how impact on CDF and/or LERF can be estimated for the RICT calculation. Include explanation of how the CRMP model is adjusted to account for post-accident containment leakage rates that are out of limits.

b. Provide justification that LCO 3.6.2 C is not a loss of function (LOF) condition in which all required trains or subsystems of a TS required system are inoperable.

Question No. 14 - LCO 3.6.6 Containment Spray Fission Product Removal Function Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of

the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

In LAR Enclosure 1, Table E1 .1 regarding Condition A of LCO 3.6.6 (Containment Spray and Cooling Systems), the disposition column states:

The PRA models the containment heat removal function consistently with the OBA [Design Basis Analysis]. The PRA does not model the fission product removal functions.

This also appears to be the case for LCO 3.6.6.B, a TS LOF condition in which both Containment Spray trains are TS inoperable because the disposition for LCO 3.6.6.A is cited in the disposition column for LCO 3.6.6.B. However, unlike Condition A, LCO 3.6.6.B is a loss of function condition which depends on use of PRA Functionality.

Questions/Discussions:

a. Explain how the Containment Spray is modeled in the PRA model (CRMP model) supporting the RICT program to reflect the design basis success criteria and how impact on CDF and/or LERF can be estimated for the RICT calculation.

b. Given that PRA does not model the fission product removal function, explain how the PRA functionality determination will account for the fission product removal function.

Question No. 15 - LCO 3.6.6, Containment Cooling System Regulatory Position 1.1.2 of RG 1.177, Revision 1 (Reference 28), states that TS requirements can be changed to reflect improved design features in a plant or to reflect equipment reliability improvements that make a previous requirement unnecessarily stringent or ineffective.

In LAR Enclosure 1, Table E1 .1, regarding LCO 3.6.6 Containment Spray and Cooling Systems, Conditions D and E, the licensee stated that the design success criteria for the containment cooling system is one of two containment cooling trains. The NRC staff notes that this design success criteria cannot be met in Condition E of LCO 3.6.6 when two containment cooling trains are inoperable, and would therefore would be considered a LOF.

The current front stop completion time for LCO 3.6.6 Condition E is 72 hours3 days <br />0.429 weeks <br />0.0986 months <br />, which exceeds the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> backstop for loss of function.

Questions/Discussions:

a. The SE for NEI 06-09 states that a RICT can be applied to SSCs that are either modeled in the PRA, or whose impact can be quantified using conservative or bounding approaches.

Explain how the Containment Cooling system is modeled in the PRA model (CRMP model) supporting the RICT program and how impact on CDF and/or LERF can be estimated for the RICT calculation consistent with NEI 06-09, Revision 0-A.

b. Justify why having two of two containment cooling trains inoperable in LCO 3.6.6 Condition Eis not considered a loss of function.

c. If LCO 3.6.6 Condition E is determined to be a LOF, confirm that SNC does not intend to take exceptions from the constraints in TS 5.5.20 for LOF conditions (items d, e, f, and h).

Question No. 16 - LCO 3.7.4 Atmospheric Relief Valves Success Criteria and LOF Regulatory Position 1.1.2 of RG 1.177, Revision 1, states that TS requirements can be changed to reflect improved design features in a plant or to reflect equipment reliability improvements that make a previous requirement unnecessarily stringent or ineffective.

LAR Enclosure 1, Table E 1.1, states regarding LCO 3. 7.4 A, Atmospheric Relief Valves (A RVs),

that there are three ARVs, however the PRA success criteria states that four of four ARVs are required for anticipated transient without trip (ATWT) conditions. The NRC staff notes there is a discrepancy between the number of ARVs covered by the LCO and the amount of ARVs required for PRA success criteria. If all of the ARVs are required to mitigate an ATWT condition, then LCO 3.7.4 Condition A, where one ARV is inoperable, and LCO 3.7.4 Condition B, where two ARVs are inoperable, would be considered a LOF.

Questions/Discussions:

a. Clarify the discrepancy in LAR Enclosure 1, Table E1 .1 regarding the number of ARVs that are covered by the LCO condition and the number of ARVs required for ATWT conditions.

b. Justify why having one or two ARVs inoperable in LCO 3.7.4 Conditions A and Bis not a LOF during ATWT conditions.

c. If LCO 3.7.4 Conditions A and/or Bare determined to be a LOF, confirm that SNC does not intend to take exceptions from the restriction in TS 5.5.20 (e) for LOF conditions (items d, e, f, and h).

Question No. 17 - LCO 3.7.6.D Condensate Storage Tank PRA Success Criteria Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

The LAR Enclosure 1, Table E1 .1, regarding the Condensate Storage Tank (CST}, proposes to credit the availability of Plant Service Water suction to the AFW pumps as the success criteria for the CST in the event the CST becomes inoperable. The disposition for LCO 3.7.6. Condition D (i.e., CST Inoperable) indicates that this condition is a TS LOF condition and states:

An NRC approval is sought as part of this LAR submittal to credit use of plant service water as modeled in the PRA as an alternate source of water to recover degraded CST design basis parameters for establishing PRA Functionality.

The statement cited above and proposed changes to LCO 3.7.6 Condition D indicate that credit for an alternate SSC other than an SSC covered by the TS is being proposed for PRA Functionality, which is inconsistent with the administrative control in TS 5.5.20, item f.1 which states:

Any structures, systems, and components (SSCs) credited in the PRA Functionality determination shall be the same SSCs relied upon to perform the specified Technical Specification safety function.

Questions/Discussions:

a. Reconcile the two statements from the LAR cited above. Include clarification of whether credit for an alternate SSC other than the SSC covered by the TS is being proposed for use in a PRA Functional determination for LCO 3.7.6 Condition D.

b. If an alternate SSC other than the SSC covered by the TS is being proposed for PRA Functionality for LCO 3.7.6 Condition D, then provide detailed justification how sufficient defense in depth and safety margins are maintained in this condition, consistent with the risk-informed principles in RG 1.174.

Question No. 18 - LCO 3.7.11.E Control Room Air Conditioning System (CRACS) PRA Success Criteria Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

LAR Enclosure 1, Table E1 .1 states regarding LCO 3.7.11.E, "Two CRACS Trains Inoperable":

Not Modeled - Documented in the PRA basis as not needed to prevent core damage.

The SE for NEI 06-09 states that a RICT can be applied to SSCs that are either modeled in the PRA, or whose impact can be quantified using conservative or bounding approaches.

Questions/Discussions:

a. If the CRACS system is not modeled in the PRA, justify why this condition can be included in the scope of the RICT program, consistent with the NEI 06-09 Revision 0-A.

b. Explain how the CRACS is modeled in the CRMP model supporting the RICT program and how a change in CDF and/or LERF can be calculated for the RICT estimate.

Include description of the CRACS success criteria that is modeled and explanation of the impact that failure of CRACS has on other systems modeled in the CRMP.

c. Clarify the difference between the design basis success criteria (I.e., one CRACS train) and the PRA success criteria (if there is any difference).

d. If the justifications requested in items a, b and c above cannot be provided, remove condition 3.7.11 E, "Two CRAGS Trains Inoperable" from the scope of the RICT program.

Question No. 19 - LCO 3.8.1.E, Two DG Sets Inoperable The LAR states that the Farley application is consistent with the Risk-Informed Technical Specifications Program approved by the NRC for SNC's Vogtle Electric Generating Plant (VEGP), Units 1 and 2 dated August 8, 2017 (Reference 29).

On page E1-13 of the LAR, the licensee states regarding LCO 3.8.1 Condition E, Two DG sets inoperable, that this is consistent with the VEGP SE. However the staff notes that the corresponding VEGP condition is marked as loss of function.

Questions/Discussions:

a. Confirm SNC's intent to treat Farley LCO 3.8.1 Condition E as loss of function.

b. Provide updated Tech Specs markups for LCO 3.8.1.E in LAR Attachments 2 and 3.

Question No. 20 - Administrative Controls Section 5.5.20 LAR Attachment 1, pages E 1-4 and E 1-5 provides the proposed addition to the TS Administrative Controls Section 5.5.20. The LAR states that this text "is consistent with TSTF505 and NEI 06-09, Revision 0-A and amended for the adjustments made to the Vogtle Electric Generating Plant (VEGP)". LAR Attachment 2, Insert 27 provides the proposed TS markup for the Administrative Controls Section 5.5.20, which appears consistent with the VEGP SE. The NRC staff notes the wording for 5.5.20.g in Attachment 1 is different than the wording for 5.5.20.g in Attachment 2, Insert 27.

Question/Discussion:

Clarify this discrepancy, and confirm that SNC intends to use the same text as the VEGP TS Administrative section, consistent with the statements made in the LAR.

Question No. 21 - Calculated RICTs that Exceed the Backstop Regulatory Position 1.1.2 of RG 1.177, Revision 1, states that TS requirements can be changed to reflect improved design features in a plant or to reflect equipment reliability improvements that make a previous requirement unnecessarily stringent or ineffective.

LAR Enclosure 1, Table E1 .2 provides example RICTs calculated for a number of LCOs (i.e.,

LCOs 3.4.11.F, Two blocks valves inoperable, 3.6.3.B and 3.6.3.C Containment Isolation Valves, 3.7.19.B, Engineered Safety Features (ESF) Room Coolers, and 3.8.7.B, Two or more required inverters inoperable) that exceed the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> backstop for conditions identified in LAR , Table E1 .1 as TS LOF conditions. The administrative control presented in TS 5.5.20 (e) stipulates that the RICT for LOF conditions may not exceed 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

Question/Discussion:

Reconcile this apparent inconsistency and confirm that SNC does not intend to take exceptions from the restriction in TS 5.5.20 (e) to not exceed the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> backstop for LOF conditions.

Question No. 22 - PRA Functionality for Systems Not Credited in the PRA NEI 06-09, Section 2.3.1, Step #11 and Section 3.2.3 provides guidance on performing PRA Functionality determination. LAR Enclosure 8, Section 2.4 discusses Farley's procedure for determining whether SSCs that are declared TS inoperable can be considered PRA functional.

According to the LAR, the procedure identifies three specific conditions in which a TS inoperable SSC can be PRA Functional. For Condition #3, the LAR states the following based on guidance from NEI 06-09:

If the condition causing the inoperability per Technical Specifications impacts only function(s) that are not modeled in CRMP and the FNP PRA has concluded that the affected function(s) has no risk impact, then the SSC may be considered PRA functional.

The LAR then provides the following example from NEI 06-09 of an inoperable SSC that fits the Condition #3:

For example, a pump backup start feature is inoperable and the feature is not credited in the PRA model (assumed failed); the RICT calculation may assume availability of the associated pump since the risk of the nonfunctional backup start feature is part of the baseline risk.

This guidance from NEI 06-09 might be misinterpreted because the definition of Condition #3 states that when the "the affected function(s) has no risk impact, then the SSC may be considered PRA functional." Yet, in the example the failure probability is assumed to be 1.0 which is reflected in the baseline risk indicating there is a risk impact.

Questions/Discussions:

a. For SSCs not modelled in the PRA that are included in the RICT program, explain which SSCs could be considered PRA Functional and which SSCs would not be considered PRA Functional. Include explanation of how different reasons for declaring an SSC as TS inoperable impacts the PRA Functionality determination.

b. Provide one or two examples (other than the example already provided) for Conditions

1. 3 that illustrate when PRA Functionality may be applied and when it may not be applied. Include explanation of why the example meets or does not meet the definition for Condition #3.

Question No. 23 - LCO 3.4.10 A Pressurizer Safety Valves Completion Time The frontstop completion time to restore a Pressurizer Safety Valve (PSV) to operable status for LCO 3.4.1 O Condition A is 15 minutes. This condition is marked as a loss of function.

Extending the completion time for Condition A by using the RICT Program will require a RICT to be calculated and PRA functionality to be determined for the inoperable PSV before the existing 15 minute frontstop is exceeded. The note for Condition A of LCO 3.4.10 also states this

condition is "not applicable when a pressurizer safety valve is intentionally made inoperable."

Question/Discussion:

Describe how the licensee will determine that an inoperable PSV can be PRA functional and calculate a RICT before the existing frontstop of 15 minutes is exceeded.

Question No. 24 - LCO 3.5.2.A ECCS Operating LAR Attachment 1 states, "Individual LCO Required Actions (RA) modified by the proposed amendment to be included in the RICT program are identified below." For LCO 3.5.2 (ECCS Operating), the LAR states that Condition A is to be modified as "One or more trains inoperable AND at least 100% of the ECCS flow equivalent to a single Operable ECCS train available."

However, in the TS markup in LAR Attachment 2 and the proposed clean-typed TS pages provided in LAR Attachment 3, Condition A states "One or more trains inoperable."

Question/Discussion:

Clarify this discrepancy between the proposed modified LCO 3.5.2 Condition A in LAR and the proposed modified LCO 3.5.2 Condition A in LAR Attachments 2 and 3.

Question No. 25 - LCO 3.8.1 EDGs and automatic load sequencers PRA modeling Given the sharing of EDGs between Unit 1 and Unit 2, discuss how the EDGs and the load sequencers are modeled in the PRA, and how a RICT is calculated for each unit when a shared Diesel becomes inoperable.

Question No. 26 - Identification of Compensatory Measures and Risk Management Actions (RMAs)

The NRC SE portion of the NEI 06-09 0-A, states that the LAR will describe the process to identify and provide compensatory measures and RMAs during extended CTs. LAR Enclosure 10 identifies four kinds of RMAs (i.e., actions to provide increased risk awareness and control, reduction of the duration of maintenance activities, reduction of the magnitude of risk increase, and minimization of the risk of a common cause failure). LAR Enclosure 10 also provides numerous specific examples of RMAs for electrical distribution related LCOs. LAR Enclosure 10 does not describe what criteria or insights (e.g., important fire areas, important operator actions) are used to determine what RMAs to apply in specific instances.

Questions/Discussions:

a. Describe what criteria or insights (e.g., important fire areas, important operator actions) are used to determine what compensatory measures and RMAs to apply in specific instances.

b. Explain how RMAs associated with a TS LOF in which PRA Functional components are credited are identified.

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ML18338A005 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DE/APLB/BC NAME SWilliams KGoldstein JCircle (ABC)

DATE 12/26/18 12/4/18 & 12/19/18 12/14/18 OFFICE NRR/DORL/LPL2-1/BC N RR/DORL/LPL2-1 /PM NAME MMarkely SWilliams DATE 1/4/19 1/4/19