ML20160A164

Supplement to the Audit Plan in Support of Review of License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times

ML20160A164
Person / Time
Site:	LaSalle
Issue date:	06/11/2020
From:	Bhalchandra Vaidya Plant Licensing Branch III
To:	Bryan Hanson Exelon Nuclear
Vaidya B
References
EPID L-2020-LLA-0018
Download: ML20160A164 (32)

Text

June 11, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LA SALLE COUNTY STATION UNIT 1 AND 2 - SUPPLEMENT TO THE AUDIT PLAN IN SUPPORT OF REVIEW OF LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT RISK-INFORMED COMPLETION TIMES (EPID L-2020-LLA-0018)

Dear Mr. Hanson:

By letter dated January 31, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20035E577), Exelon Generation Company, LLC (Exelon, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC) amend the technical specifications (TSs) (Appendix A) of Renewed Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station, Units 1 and 2 (LaSalle).

Exelons proposed license amendment request (LAR) would revise TS requirements to permit the use of risk-informed completion times for actions to be taken when limiting conditions for operation are not met. The proposed changes are based on Technical Specifications Task Force Traveler (TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times -

RITSTF Initiative 4b, dated July 2, 2018 (ADAMS Package Accession No. ML18269A041).

The NRC staff has reviewed Exelons LAR and determined that a regulatory audit would assist in the timely completion of the LAR review. The initial audit plan consisting of In-office Audit and Site Audit was provided to the licensee on March 30, 2020 (ADAMS Accession No. ML20090F585). This letter provides an update to the audit plan to reflect a remote audit, and the detailed audit agenda along with questions the staff has prepared to be discuss during the remote audit. The NRC staff will conduct a regulatory audit to support its review of the LAR in accordance with the enclosed audit plan.

The audit will be conducted from June 15, 2020, to June 19, 2020, remotely.

It should be noted that the audit for this LAR and regulatory audit for the risk-informed categorization and treatment of structures, systems, and components LAR are being conducted concurrently. The logistics and scope of the audit supplement were discussed with your staff on June 10, 2020. The audit plan supplement is enclosed.

B. Hanson If you have any questions, please contact me by telephone at 301-415-3308 or by e-mail to Bhalchandra.Vaidya@nrc.gov.

Sincerely,

/RA/

Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosure:

Audit Plan Supplement cc: Listserv

AUDIT PLAN SUPPLEMENT REGARDING RISK-INFORMED COMPLETION TIMES EXELON GENERATION COMPANY, LLC LA SALLE COUNTY STATION UNITS 1 AND 2 DOCKET NO. 50-373 AND 50-374

1.0 BACKGROUND

By letter dated January 31, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20035E577), Exelon Generation Company, LLC (Exelon, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC) amend the technical specifications (TSs) (Appendix A) of Renewed Facility Operating License Nos. NPF-11 and NPF-18 for La Salle County Station, Units 1 and 2 (LaSalle). Exelons proposed license amendment request (LAR) would revise TS requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation are not met. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler (TSTF)-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b, dated July 2, 2018 (ADAMS Package Accession No. ML18183A493).

2.0 REGULATORY AUDIT BASES A regulatory audit is a planned license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. The audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. Performing a regulatory audit is expected to assist the NRC staff in efficiently conducting its review and gaining insights for licensees processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket.

The basis of this audit is the licensees LAR for LaSalle, and the Standard Review Plan Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (ADAMS Accession No. ML071700658).

The audit will be performed consistent with NRC Office Instruction LIC-111, Revision 1, Regulatory Audits, dated October 31, 2019 (ADAMS Accession No. ML19226A274). An audit was determined to be the most efficient approach toward a timely resolution of issues associated with this LAR review, since the staff will have an opportunity to minimize the potential for multiple rounds of requests for additional information (RAIs) and ensure no unnecessary burden will be imposed by requiring the licensee to address issues that are no longer necessary to make a safety determination.

Enclosure

3.0 PURPOSE AND SCOPE The purpose of this audit is to:

Gain a better understanding of the calculations, analyses, and bases underlying the LARs. Confirm the staffs understanding of the LARs.

Gain a better understanding of the approach for developing and implementing nuclear power station risk-managed TS programs.

Identify information that the licensee should submit for NRC staff to reach a regulatory decision. Discuss potential RAIs.

Gain a better understanding of the extent that the licensees proposed amendment to modify TS requirements for RICTs is consistent with TSTF-505, Revision 2, and Nuclear Energy Institute (NEI) 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, dated November 6, 2006.

Gain a better understanding of 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.

Gain a better understanding of the technical acceptability of the probabilistic risk assessment (PRA) for use in the application and how plant design features are modeled in the PRA used to support the LAR.

The areas of focus for the regulatory audit are the information contained in the LAR, the audit information needs are listed in the following section of this audit plan supplement, and all associated and relevant supporting documentation (e.g., methodology, process information, calculations, etc.). The relevant supporting documents are identified below.

4.0 INFORMATION AND OTHER MATERIAL NECESSARY FOR THE REGULATORY AUDIT The following documentation should be available to the audit team:

1. Reports of peer reviews (full-scope and focused-scope), self-assessments, and facts and observations (F&Os) closure reviews for the internal events, internal flooding, and FPRAs cited in LaSalles LAR dated January 31, 2020;

2. Uncertainty notebooks for the LaSalle internal events, internal flooding, and FPRAs related to PRA model assumptions and sources of uncertainty;

3. Documentation on evaluation of the generic and plant specific uncertainties with respect to the Lasalle LAR dated January 31, 2020;

4. Documentation supporting the example RICT calculations presented in LAR Enclosure 1 Table E1-2, including determination of corresponding risk management actions;

5. PRA notebooks for the modeling of FLEX equipment and FLEX human error probabilities, if credited in the PRA;

6. Results of the FPRA and resolution of F&Os;

7. Draft or final RICT program procedures (e.g., procedures related to risk management actions, PRA functionality determination, and recording limiting conditions for operation);

8. The loss of function (LOF) analysis and the licensee models for instrumentation and control (I&C) in PRA. Also, any documents that support the defense-in-depth justification; and

9. We need to understand the interactions between the electrical divisions AC (alternate current) and direct current (DC) that that require an opposite unit division system to be operable. If there are shared support systems (essential cooling water, heat ventilation, and air conditioning (HVAC), etc.), we would like to understand the systems involved. For the diesel generators (DGs), what are the worst-case scenarios for the loss of multiple DGs from different divisions for a potential accident on one unit. Under what scenarios, are the HPCS Division 3 DGs for each unit required? Are the five DGs for both units considered a system where only one failure may be assumed?

Applicable documents:

Single line diagrams of the electrical systems, with description of:

a) normal alignment, b) any limiting condition for operation-related configuration if one or more power source(s) is/are inoperable including interconnection of DGs which may require DC power for associated breakers.

c) any unique accident mitigation alignment credited in the failure mode analyses for calculating RICT d) Station Blackout considerations e) indicate where loss of function occurs for loss of multiple power sources, system descriptions if available for DGs and DC system

10. For loss of offsite power (both circuits) a 30-day RICT is proposed.

a. We need to understand the design of the cooling water system or ultimate heat sink (UHS) so that the DGs, residual heat removal service water(RHRSW), etc.,

can run for 30 days with cooling available. The licensing basis for UHS is a 30-day inventory. A preliminary review of the UHS description in updated final safety analysis report (UFSAR) seems to indicate that nonsafety-related systems are not required to maintain cooling for ECCS and DGs.

b. Some plants need station air or instrument air for cycling valves. We need to understand if air systems for critical valves (containment isolation, residual heat removal etc.) need nonsafety-related power sources to maintain systems in operable condition over a 30-day period.

Applicable documents:

System description and flow diagrams associated with UHS and DG cooling systems and if applicable station air/instrument air systems.

11. TS 3.3.8.1 Loss-of-Power (LOP) Instrumentation: we need to understand the success criterion and what systems (DG and offsite power) are assumed to fail when this LCO is extended. Enclosure 1, Page E1-7, assumes only the DG fails to start. LCO 3.3.8.1 A is related to One or more channels inoperable. We need to understand the interpretation of

or more as related to a single division and as related to redundant divisions, as there are several channels associated with LOP instrumentation Applicable Documents:

Description of LOP Instrumentation and applicable failure modes considered in the RICT

12. Any other supporting documentation that the licensee may determine is responsive to the NRC staffs above information requests.

5.0 AUDIT TEAM The members of the audit team are anticipated to be:

Adrienne Brown, Reliability and Risk Analyst, PRA, NRC Todd Hilsmeier, Reliability and Risk Analyst, PRA, NRC Shilp Vasavada, Reliability and Risk Analyst, PRA, NRC De Wu, Reliability and Risk Analyst, PRA, NRC Stacey Rosenberg, Branch Chief, PRA Licensing Branch C, NRC Robert Pascarelli, Branch Chief, PRA Licensing Branch A, NRC Robert Vettori, Fire Protection Engineer, NRC Bhalchandra Vaidya, Project Manager, NRC Garill Coles, Principal Engineer, Pacific Northwest National Laboratory (NRC Contractor)

Mark Wilk, Pacific Northwest National Laboratory (NRC Contractor)

John Bozga, Region III John Honcharik NRC G. Bedi, NRC Carte Norbert, NRC Andrea Russell, NRC Victor Cusumano, Branch Chief, Technical Specifications Branch, NRC Joseph Ashcraft, NRC Matharu Gurcharan, NRC Ed Kleeh, NRC 6.0 LOGISTICS The audit will be conducted from June 15, 2020, to June 19, 2020, remotely, between 8:30 a.m. and 4:00 p.m. each day. An entrance briefing will be held at the beginning of the first part of the audit, and an exit briefing will be held at the end of the second part of the audit. contains the agenda for the remote audit and identified details of the topics the staff would like to have prepared dialogue. Attachment 2 contains Audit Questions. The NRC project manager will coordinate any identified changes to the audit schedule and logistics with the licensee.

7.0 SPECIAL REQUESTS The NRC staff would like access to the documents listed above in Section IV through an online portal that allows the NRC staff and contractors to access documents via the internet. The following conditions associated with the online portal must be maintained throughout the

duration that the NRC staff and contractors have access to the online portal:

The online portal will be password-protected, and separate passwords will be assigned to the NRC staff and contractors who are participating in the audit.

The online portal will be sufficiently secure to prevent the NRC staff and contractors from printing, saving, downloading, or collecting any information on the online portal.

Conditions of use of the online portal will be displayed on the login screen and will require acknowledgement by each user.

User name and password information should be provided directly to the NRC staff and contractors. The NRC project manager will provide Exelon the names and contact information of the NRC staff and contractors who will be participating in the audit. All other communications should be coordinated through the NRC project manager.

Visitor access for the plant walkdown to audit aspects of the proposed alternate seismic approach. Alternately, videos and/or photos, to support a virtual walkdown, as a contingency.

Access to licensee and licensees contractor personnel knowledgeable in the proposed alternate seismic approach for categorization, plant design, operation and any supporting PRA(s) used to address the staffs audit questions.

8.0 DELIVERABLES An audit summary, which may be public, will be prepared within 90 days of the completion of the audit. If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the NRC staff will issue RAIs to the licensee after the audit.

Audit Agenda Day 1 - Monday, June 15, 2020 (8:30 a.m. to 4:00 p.m.)

Morning Kick-off. Opening comments - NRC and Exelon. Introductions and logistics.

Real-time risk demonstration by Exelon.

Discussion on preparing and benchmarking the real-time risk model, including how seasonal variations are accounted (refer to Attachment 2).

Discussion on internal events probabilistic risk assessment (PRA) technical acceptability for TSTF-505, 10 CFR 50.69.

TSTF-505, APLA Audit Questions 01, 04, 05, LUNCH: 12:00 p.m. - 1:00 p.m.

Afternoon Modeling of PRA systems and use of surrogates (APLA Audit Question 07).

Calculation of risk-informed completion time estimates.

Electrical and I&C (Audit Questions) -2 hours starting at 2:00 pm EST.

Summary of the day.

NRC staff meeting.

Day 2 - Tuesday, June 16, 2020 (8:30 a.m. to 4:00 p.m.)

Morning Summary of previous day.

Key assumptions and key sources of uncertainty (TSTF-505, APLA Audit Questions 09, 10, and 11, 10 CFR 50.69, APLA Audit Question 04, refer to Attachment 2).

Modeling of instrumentation and controls in the PRA, and loss of function (TSTF-505, APLA Audit Questions 06 and 08, refer to Attachment 1).

Electrical and I&C (Follow up to Audit Questions) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> starting at 11:00 am EST LUNCH: 12:00 p.m. - 1:00 p.m.

Afternoon FLEX credit (10 CFR 50.69, APLA Audit Question 08).

FPRA technical acceptability (TSTF-505, APLA Audit Questions 02 and 03, refer to Attachment 2).

Summary of the day.

NRC staff meeting.

Day 3 - Wednesday, June 17, 2020 (8:30 a.m. to 4:00 p.m.)

Morning Summary of previous day.

Electrical and I&C (Follow up to Audit Questions) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> starting at 11:00 am EST Key Principle 5, Maintenance Rule and monitoring (TSTF-505, APLA Audit Question 12).

10 CFR 50.69 Process Overview:

Attachment 1

10 CFR 50.69, APLA Audit Questions 06, 07, and 10.

LUNCH: 12:00 p.m. - 1:00 p.m.

Afternoon External hazards discussion for TSTF-505 (APLC Audit Questions 01)..

NRC staff meeting.

Summary and exit of TSTF-505 audit.

Partial summary of 10 CFR 50.69 audit (except alternate seismic approach and external hazards).

Day 4 - Thursday, June 18, 2020 (8:30 a.m. to 4:00 p.m.)

Morning and Afternoon Alternate seismic approach for 10 CFR 50.69 (APLC Audit Questions 01 - 13).

Day 5 - Friday, June 19, 2020 (8:30 a.m. to 4:00 p.m.)

Morning Alternate seismic approach for 10 CFR 50.69 (APLC Audit Questions 01 - 13).

Afternoon Alternate seismic approach for 10 CFR 50.69 (APLC Audit Questions 01 - 13).

External hazards discussion for 10 CFR 50.69 (APLC Audit Questions 14).

Summary and exit of 10 CFR 50.69 audit.

Details for Agenda Topics of Discussion Day 1 (Morning): Calculations for Seasonal Variations LAR Enclosure 8, Section 2 states, [t]herefore, no adjustment to initiating event frequencies in required within the RTR [Real Time Risk] tool.

Given that initiating event frequency data could include data that potentially changes during the year, address the following:

What seasonal dependencies are accounted in the RTR and how are they modeled?

Do any of the initiating event data used in the PRA models include data that is sensitive to seasonal variations?

Discuss how changes in initiator frequency due to seasonal variations is accounted for in the RTR model used in the RICT calculations.

Day 2 (Morning):

(1) Assumptions and sources of uncertainty (TSTF-505, APLA Audit Questions 09 and 10):

Discussion of process used (i.e., NUREG 1855, Revision 1),

List of assumptions and sources of uncertainty identified for base PRA, List of key assumptions and sources of uncertainty identify for TSTF-505 application, Dispositions/treatment:

o Cable selection, o Vapor suppression capability, o Hardened containment vent, o Identification of target sets within zone of influence (ZOI),

o Survivability of ECCS after containment vent failure, o Digital feedwater control probabilities (10 CFR 50.69, APLA Audit Question 04).

(2) State-of-Knowledge Correlation (TSTF-505, APLA Audit Question 11).

Day 2 (Morning):

(1) Instrumentation and Control Modeling: Loss of Function (TSTF-505, APLA Audit Question 08):

TS LCO 3.3.1.1 (Reactor Protection System (RPS) Instrumentation) Condition A (one or more required channels inoperable).

TS LCO 3.3.1.1 (RPS Instrumentation) Condition B (One or more functions with one or more required channels inoperable in both trip systems).

TS LCO 3.3.2.2 (Feedwater System and Main Turbine High Water Level Trip Instrumentation) Condition A (One or more channels inoperable).

TS LCO 3.3.4.2 (Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation) Condition A (one or more channels inoperable).

TS LCO 3.3.6.1 (Primary Containment System Isolation Instrumentation)

Condition A (One or more channels inoperable).

Attachment 1

TS LCO 3.3.8.1 (Loss of Power (LOP) Instrumentation) Condition A (One or more channels inoperable).

TS LCO 3.5.1 (ECCS - Operating) Condition B (High Pressure Core Spray (HPCS) System inoperable TS LCO 3.5.3 (RCIC System) Condition A (RCIC System inoperable).

TS LCO 3.6.1.3 (Primary Containment Isolation Valves (PSIV)) Condition A (One or more penetration flow paths with one PCIV inoperable for reasons other than Condition D).

TS LCO 3.8.7 (Distribution Systems - Operating) Condition A (One or both Division 1 and 2 AC electrical power distribution subsystem inoperable).

TS LCO 3.8.7 (Distribution Systems - Operating) Condition B (One or both Division 1 and 2 125 VDC electrical power distribution subsystem inoperable)

TS LCO 3.8.7 (Distribution Systems - Operating) Condition D (One or more required opposite unit Division 2 AC or DC electrical power distribution subsystem inoperable).

Day 2 (Afternoon): FPRA Technical Acceptability (1) Disposition of open facts and observations (F&Os) (TSTF-505, APLA Audit Question 02).

(2) FPRA methods (TSTF-505, APLA Audit Question 03): NRC staff would like licensee personnel to be prepared to discuss the identified FPRA methodologies below to validate and verify only NRC-accepted methods have been incorporated into the PRA.

Please have the applicable FPRA notebooks and logic models available during the discussion.

o Use of unacceptable methods, Identification of any method that deviate from NUREG/CR-6850 o Reduced transient heat release rates, Identification of fire areas where a reduced transient fire heat release rate (HRR) is credited and the reduced HRR value that was applied Discuss any administrative controls that support justification for using the reduced HRR discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance Discussion of personnel traffic that would be expected through each location discuss the results of any review of records related to compliance with the transient combustible and hot work controls o Minimum joint human error probability, Identification of any minimum joint human error probability (HEP) values less than 1E-05 used in the FPRA o Obstructed plume model (NUREG-2178),

Discuss if the base of the fire was assumed to be located at an elevation of less than one half of the cabinet o Well-sealed motor control center (MCC) cabinets (Frequently Asked Question (FAQ) 08-0042),

Discuss how fire propagation outside of the well-sealed MCC cabinets

greater than 440 V is evaluated Identify if any well-sealed cabinets less than 440 V are included in Bin 15 count of ignition sources o Influence factors for transient fires (FAQ 12-0064),

Confirm the influencing factors for hot work and transient fire frequencies applied (e.g., NUREG/CR-6850 or FAQ 12-0064)

Identify and discuss any combustible control violations and treatment of these violations for the assignment of the transient fire frequency influence factors Identify any physical analysis units (PAUs) assigned an influencing factor of 0 for maintenance, occupancy, storage, or hot work o Treatment of fire dependencies between units, Discuss how the risk contributions of fires originating in one unit is addressed for the other unit Discuss how the contribution form fires in common areas are addressed, including the risk contribution of fires that can impact components in both units o Overall peer review of fire methods, o Very early warning fire detection systems (NUREG-2180).

AUDIT QUESTIONS LASALLE COUNTY STATION, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B DOCKET NOS. 50-373 AND 50-374 RENEWED FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18

Background:

By application dated January 31, 2020, Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) for LaSalle County Station, Units 1 and 2 (LaSalle) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20035E577). The amendment would revise technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) are not met. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated July 2, 2018 (ADAMS Accession No. ML18183A493). The U.S. Nuclear Regulatory Commission (NRC) has determined that the following information is needed in order to complete its review.

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML17317A256), states that the scope, level of detail, and technical adequacy of the probabilistic risk assessment (PRA) are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process. The NRCs safety evaluation (SE) for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b: Risk Managed Technical Specification (RMTS) (ADAMS Accession Nos. ML071200238 and ML122860402), states that the PRA models should conform to the guidance in RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. The current version is RG 1.200, Revision 2 (ADAMS Accession No. ML090410014), which clarifies the current applicable American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard is 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. RG 1.200 describes a peer review process using ASME/ANS RA-Sa-2009 as one acceptable approach for determining the technical acceptability of the PRA. The primary result of a peer review are the facts and observations (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 Nuclear Energy Institute (NEI) guidance documents NEI 05-04, NEI 07-12, and NEI 12- 13, titled NEI 05-04/07-12/12-[13] Appendix X: Close-out of Facts and Observations (F&Os) (ADAMS Package Accession No. ML17086A431), which was accepted by the NRC in a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427).

Attachment 2

APLA QUESTION 01 - Internal Events PRA Self-Assessment Findings LAR Enclosure 2, Section 3, states that the last full-scope internal events peer review was performed in April 2008 using the ASME RA-Sc-2007 PRA standard, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum C to RA-S-2002, dated August 2007, and RG 1.200, Revision 1 (ADAMS Accession No. ML070240001). The LAR explains that a gap analysis to the ASME/ANS RA-Sa-2009 PRA Standard and RG 1.200, Revision 2 (ADAMS Accession No. ML090410014), was performed as part of the 2014 LSCS PRA update. The LAR states that the gaps were mostly related to unresolved F&Os at the time of the update and that these gaps are identified in LAR Enclosure 2, Table E2-1. The NRC staff observes that three gaps were listed in the last entry in LAR Table E2-1 for internal flooding PRA associated with supporting requirements (SRs) IFSO-A3, IFSN-A7, and IFQU-A3, and the complete wording of the gap assessment findings, basis for significance, and possible resolutions are not provided in LAR Table E2-1. The LAR dispositions for these gaps state that these are documentation issues that will be resolved in the next LaSalle PRA update. It is unclear to the NRC staff that the cited SRs are only documentation issues as it pertains to fulfillment of the SRs. Additionally, as part of its audit of the licensees LAR (ADAMS Accession No. ML20090F585), the NRC staff reviewed the 2014 gap assessment and observed that additional SRs were not met or were met at Capability Category (CC) I (e.g., DA-C6, DA-C7, DA-C8, DA-C10, SC-A5, and HR-D3) and not evaluated in LAR Table E2-1. It is not clear whether these gaps are still unresolved in the PRA models and whether any of these unresolved gaps can impact the RICT calculations. To confirm that these unresolved gaps and any additional SRs determined to not meet CC II have no impact on the RICT calculations address the following:

a) Provide the complete wording of the gap assessment findings, basis for significance, and possible resolutions (to the extent this information exists) against SRs IFSO-A3, IFSN-A7, and IFQU-A3 to support the conclusion that the resolution of these gap findings represent documentation issues that cannot impact the RICT calculations.

Include a discussion on whether any updated documentation (or documentation that will be updated) reflect a new analysis or material that is not yet peer reviewed.

b) Explain whether any additional gaps identified from the 2014 gap assessment that were not included in LAR Table E2-1 (e.g., those gaps associated with SRs DA-C6, DA-C7, DA-C8, DA-C10, SC-A5, and HR-D3) remain unresolved in the PRA models of record. If any gaps against the PRA standard SRs remain unresolved in the PRA models of record, then:

i. Identify the gaps that remain unresolved in the PRA models of record and justify that they have no impact on the risk-informed application (e.g., RICT calculations).

ii. If justification for these gaps cannot be provided in part (b).i above, then propose a mechanism that ensures these gaps are resolved prior to implementation of the RICT program (e.g., include as an implementation item in Attachment 5 of the LAR).

APLA QUESTION 02 - Open FPRA Facts and Observations (F&O)

LAR Enclosure 2, Table E2-2 presents the dispositions for three F&Os that remain open after the Independent Assessment (IA) performed for closure of F&Os; two that remain open (i.e., 1-19, 4-17) and one partially resolved (i.e., 6-11). For F&O 1-19, the finding has been resolved but the resolution has not yet been reviewed by the IA team. For F&Os 4-17 and 6-11, the licensee states the items will be resolved prior to TSTF-505 implementation, however, of the LAR that includes a table of implementation items to be completed prior to implementation of the RICT program does not include these items. In light of these observations, provide the following:

a) Regarding F&O 1-19, the LAR disposition acknowledges that the update was not reviewed by the IA team, and does not discuss the results of the systemic review.

i. Describe the results from the systemic review of the circuit evaluation package notes and assumptions and explain what FPRA (FPRA) modelling adjustments have been determined to be needed.

ii. Describe the modelling updates that have been incorporated into the PRAs demonstrating that the identified modelling concerns are addressed.

iii. Alternatively, propose a mechanism that ensures the results of the systemic review and the updates to the FPRA are reviewed by the IA team and the F&O is closed prior to implementation of the RICT program (e.g., include as an implementation item in Attachment 5 of the LAR).

b) Regarding F&O 4-17, the LAR disposition states that [t]his item will be resolved prior to TSTF-505 implementation. Furthermore, the licensee states that the impact of this issue is judged to be minimal. However, it is not clear to NRC staff the impact on the RICT calculations. Therefore, address the following:

i. Provide justification (e.g., description and results of a sensitivity study) that any needed adjustments made to the FPRA identified from review of the plant-specific data on the fire suppression and detection systems cannot impact the RICT calculations.

ii. Alternatively, propose a mechanism that ensures the review of plant-specific data for fire suppression and detection systems is performed and any update needed to the FPRA is completed prior to implementation of the RICT program (e.g., include as an implementation item in Attachment 5 of the LAR).

c) Regarding F&O 6-11, the LAR disposition discusses that a review will be performed to verify consistency with NEI 00-01, Revision 3, prior to implementation of the RICT program. However, no commitment to complete an implementation item for this F&O is made in LAR Attachment 5. Also, the NRC staff notes that in the event the review cannot verify the circuit analysis was performed in accordance with the requirements of NEI-00-01, Revision 3, then adjustments to the FPRA model may be needed.

i. Provide sufficient justification to support the conclusion provided in table E2-2 that any revisions to the cable selection based upon historical methods used

during the time of the analysis development and more recent guidance (i.e.,

NEI 00-01, Revision 3) has no impact on the TSTF-505 RICT calculations.

ii. Alternatively, propose a mechanism that ensures the review of the circuit analysis to the requirements of NEI-00-01, Revision 3 is performed and any needed update to the FPRA is completed prior to implementation of the RICT program (e.g., include as an implementation item in Attachment 5 of the LAR).

APLA QUESTION 03 - Fire Hazards for LaSalle TSTF-505 RG 1.200 states NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. Some concerns are not always readily identified in F&Os by the peer review teams but are considered potential key assumptions by the NRC staff because using more defensible and less simplified assumptions could substantively affect the fire risk and overall risk profile of the plant. The NRC staff notes that the calculated results of the PRA are used directly to calculate a RICT, which subsequently determines how long systems, structures, and components (SSCs) (both individual SSCs and multiple, unrelated SSCs) controlled by TSs can remain inoperable. The PRA results are given a very high weight in a TSTF-505 application; therefore, the NRC staff requests additional information on the following areas that have been identified as potentially key FPRA assumptions.

LAR Enclosure 9, Section 4, states that the LaSalle FPRA was developed using consensus methods outlined in NUREG/CR-6850 and interpretations of technical approaches as required by NRC. The licensee further states the FPRA methods were based on , other more recent NUREGs, (e.g., NUREG-7150 []), and published frequently asked questions (FAQs) for the FPRA. Furthermore, Part (e) of the proposed TS 5.5.17 ("Risk Informed Completion Time Program) states in part, [m]ethods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

There have been changes to the FPRA methodology since the last full-scope peer review of the LaSalle FPRA in 2015. The integration of NRC-accepted FPRA methods and studies described below that are relevant to this submittal could potentially impact the TSTF-505 RICT calculations and/or risk metrics for total core damage frequency (CDF) and total large early release frequency (LERF):

NUREG-2178, Volume 1, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), dated April 2016 (ADAMS Accession No. ML16110A140).

NUREG-2180, Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE), dated December 2016 (ADAMS Accession No. ML16343A058).

Section 2.5.5 of RG 1.174 provides guidance that indicates additional analysis is necessary to ensure that contributions from the above influences would not change the conclusions of the LAR.

For each of the above NRC-accepted FPRA methods and studies, the NRC staff requests the licensee address one of the following:

a) Discuss how the FPRA method/study had been incorporated into the LaSalle FPRA and, as applicable, summarize the changes made to the FPRA model. Indicate whether this change was PRA maintenance or a PRA upgrade as defined in ASME/ANS RA-Sa-2009, Section 1-5.4, as qualified by RG 1.200, Revision 2, along with a justification for the determination. If this change constitutes a PRA upgrade, discuss the focused-scope (or full-scope) peer review(s) that has been performed to evaluate the change, and provide any open F&Os and associated dispositions from this peer review(s) in accordance with RG 1.200, Revision 2.

OR b) If the FPRA method/study has not been incorporated into the LaSalle FPRA, provide a detailed justification for why the integration of the FPRA method/study would not change the conclusions of the LAR, and subsequently not impact the TSTF-505 RICT calculations and/or risk metrics for total CDF and total LERF. As part of this justification, identify any FPRA methodologies used in the LaSalle FPRA that are no longer accepted by the NRC staff (e.g., guidance provided in frequently asked question (FAQ) 08-0046, Closure of National Fire Protection Association 805 Frequently Asked Question 08-0046 Incipient Fire Detection Systems, ADAMS Accession No. ML093220426, has been retired by letter dated July 1, 2016, (ADAMS Accession No. ML16167A444)). Provide technical justification for its use and evaluate the significance of its use on the risk metrics for this application provided in Enclosure 5 of the LAR.

OR c) Propose a mechanism that ensures the FPRA method/study (or other NRC acceptable method) will be integrated into the LaSalle PRA prior to implementation of the RICT program. If this FPRA update is determined to be a PRA model upgrade per the ASME/ANS PRA Standard, include in this mechanism a process for conducting a focused-scope peer review and ensure any findings are closed by using an approved NRC process.

Please ensure that the following specific topic areas are addressed in the response: minimum joint human error probability, obstructed plume model, well-sealed MCC cabinets, influence factors for transient fires, reduced transient heat release rate, and PRA treatment of fire dependencies between Units 1 and 2.

APLA QUESTION 04 - PRA Model Update Process Section 2.3.4 of NEI 06-09, Revision 0-A, specifies that [c]riteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.

LAR Enclosure 7 states that if a plant change or a discovered condition is identified and can have significant impact to the RICT program calculations then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the change will be

incorporated into applicable PRA models without waiting for the next periodic PRA update. The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate the update.

Therefore, describe the conditions under which an unscheduled PRA update (i.e., more than once every two refueling cycles) would be performed and the criteria that would be used to require a PRA update. In the response define what is meant by significant impact to the RICT Program calculations.

APLA QUESTION 05 - TSTF-505 - Consideration of Shared Systems in RICT Calculations RG 1.174, Revision 3, states that the scope, level of detail, and technical adequacy of the PRA are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process.

LAR Enclosure 12 refers to the use of cross-ties in response to loss of offsite power and station blackout events. Furthermore, LAR Enclosure 2, LCO 3.8.7.d states [t]he portions of the opposite unit's Division 2 AC [alternating current] and 125 V DC [volt direct current] electrical power distribution subsystems capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3. The LCO and identified operator action in Enclosure 12 of the LAR infers that the function of certain systems can be shared across units. It is not clear to NRC staff what systems are capable of supporting both units and how the shared systems are credited for both units in the PRA models. NRC staff notes that for certain events such as dual unit events (e.g.,

loss of offsite power) the shared systems can only be credited for one unit. In light of these observations, provide the following information:

a) Confirm whether shared systems are credited for both units in the Real Time Risk (RTR) model that supports the RICT calculations, if so, provide a list of those systems.

b) If shared systems are credited in the RTR model that supports the RICT calculations, then describe how each shared system is modelled for each unit in dual unit events that can create a concurrent demand for the shared system.

c) If a shared system is credited in the RTR model that supports the RICT calculations and the impact of dual unit events is not addressed in the RTR model, then justify that this exclusion does not impact the RICT calculations.

APLA QUESTION 06 - TSTF 505 - Instrumentation and Controls (I&Cs)

Concerning the quality of the PRA model, NEI 06-09, Revision 0-A, states that RG 1.174 and RG 1.200 define the quality of the PRA in terms of its scope, level of detail, and technical acceptability. The quality of the PRA must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.

Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties be considered in application of the PRA base model results to the RICT program. The NRC SE for NEI 06-09, Revision 0, states that this consideration is consistent with Section 2.3.5 of RG 1.177, Revision 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications" (ADAMS Accession No. ML100910008). NEI 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results

of a RICT calculation and that sensitivity studies should be used to develop appropriate compensatory risk management actions (RMAs). The following additional information is requested:

a) For several TS LCO conditions listed in LAR Table E1-1, the table indicates that the detailed I&C modeling is not included in the PRA models to reflect the associated TS LCO, therefore the inoperability of the SSC (e.g., channel) will be modeled using a surrogate event. For other TS LCOs in the RICT program, it is not clear to the NRC staff when the I&C modeling is included in the PRA in sufficient detail to support implementation of TSTF-505, Revision 2.

Describe how I&C SSCs that are applicable for the RICT program are modeled/considered in the PRA. Include in this discussion: (1) the scope of the I&C equipment that is explicitly modeled (e.g., bistables, relays, sensors, integrated circuit cards), (2) description of the level of detail that the PRA model supports (e.g., are all channels of an actuation circuit considered), (3) discussion of the generic data and plant-specific data used, and (4) discussion of the associated TS functions for which a RICT can be applied.

b) Regarding digital I&C, the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist such as lack of industry data for digital I&C components, differences between digital and analog system failure modes, and the complexities associated with modeling software failures including common cause software failures. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program.

In light of these observations, identify which LCOs are determined to be impacted by digital I&C system modeling for which RMAs will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation require additional RMAs.

APLA QUESTION 07 - System and Surrogate Modeling Used in the PRA Models The NRC SE to NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions and that justification be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions. Table E1-1 in Enclosure 1 to the LAR identifies each TS LCO proposed to be included in the RICT program and describes how the systems and components covered in the TS LCO are implicitly or explicitly modeled in the PRA. For some TS LCO conditions, the table explains that the associated SSCs are not modelled in the PRAs but will be conservatively represented using a surrogate event. For certain LCOs, it is unclear to the NRC staff whether the modeling (i.e., depth, surrogates used) will be acceptable to address the associated RICTs calculated. Therefore, address the following:

a) For TS LCO 3.3.6.1 (Primary Containment Isolation Instrumentation) Condition A (Primary containment instrumentation - one or more channels inoperable) in LAR Table E1-1, the logic for primary containment isolation is not modeled in detail and, therefore, a surrogate event will be used. The table states that the surrogate event will either be failure of containment or failure of the frontline system. It is not clear to NRC staff what system is being referred to by the phrase frontline system. Confirm/identify the

frontline system(s) intended to be referred to in the table (e.g., primary containment isolation valves (PSIV)) and discuss how the failure of the function(s) compares to failure of containment.

b) For TS LCO 3.6.1.3 PSIV, Condition A (one or more penetration flow paths with one PCIV inoperable for reasons other than Condition D), in LAR Table E1-1, it states not all PSIVs are modeled, therefore, a surrogate of a pre-existing containment failure is chosen. It is not clear to the NRC staff what pre-existing containment failure will be chosen and how LaSalle will ensure that it represents the equivalent release as an open PSIV. Therefore, discuss the pre-existing containment failure or criteria used to determine which surrogate will be identified for an inoperable PSIV and provide sufficient justification to ensure that the surrogate will reflect the function covered by the TS LCO condition. In the discussion include whether this failure assumes a large early release.

APLA QUESTION 08 - TSTF-505 - Potential Loss of Function Conditions TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493) does not allow for TS loss of function conditions (i.e., those conditions that represent a loss of a specified safety function or inoperability of all required trains of a system required to be operable) in the RICT program.

LAR Enclosure 1, Table E1-1, includes a list of the revised require actions proposed for the LSCS RICT program. It is not clear to the NRC staff how each TS LCO was assessed to determine if the condition could result in a TS loss of function as described in TSTF-505, Revision 2.

Therefore, address the following:

a) Provide criteria used to assess how each LCO does not represent a TS loss of function.

b) If in the response to item (a) above it cannot be justified that one or more of the LCOs does not represent a TS loss of function, then remove the LCO from the RICT program and provide an updated TS markup.

APLA QUESTION 09 - Process: PRA Model Uncertainty Analysis The NRC staff SE to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and sources of uncertainty and assess and disposition each as to their impact on the RMTS application. Section 5.3 of NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Main Report," dated March 2017 (ADAMS Accession No. ML17062A466), presents guidance on the process of identifying, characterizing, and qualitatively screening model uncertainties.

LAR Enclosure 9 states that the process for identifying key assumptions and sources of uncertainty for the internal events and FPRAs was performed using the guidance in NUREG-1855, Revision 1. The LAR indicates that in addition to reviewing generic industry sources of uncertainty for applicability, the internal events PRA (IEPRA) and FPRA models and notebooks were reviewed for plant-specific assumptions and sources of uncertainty. As part of the NRC staffs audit (ADAMs Accession No. ML20090F585) of the licensees LAR, the NRC staff reviewed the PRA assumptions and sources of uncertainty associated with the IEPRA (includes internal floods) and FPRA models. Based upon the information provided in the LAR, it is not clear to the NRC staff how the licensee considered the generic industry sources of uncertainty provided in EPRI report 1026511 for the base IEPRA. Also, it is not clear to the NRC staff what

evaluation criteria was used to consistently screen uncertainties from an initial comprehensive list of generic, plant-specific, and modelling choice uncertainties in order to produce the list of key assumptions and sources of uncertainty that are presented in LAR Enclosure 9.

In light of these observations, provide a detailed summary of the process used to identify the assumptions and sources of uncertainty presented in the LAR. The discussion should include:

a) How the process is consistent with NUREG-1855, Revision 1, or other NRC-accepted methods. If deviating from the current guidance provided in NUREG-1855, Revision 1, provide a basis to justify the appropriateness of any deviations for use in the risk-informed application (e.g., exclusion/consideration of EPRI report1026511).

b) A description of how the key assumptions and sources of uncertainties provided in the LAR were identified from the initial comprehensive list of PRA model (i.e., base PRA models) uncertainties and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns.

APLA QUESTION 10 - Treatment: PRA Model Uncertainty Analysis The NRC staff SE to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and key sources of uncertainty and assess/disposition each as to their impact on the RMTS application. LAR Enclosure 9, Tables E9-1, E9-2, and E9-3, provided the key assumptions and sources of uncertainty identified for the IEPRA, transition to the RTR model, and the FPRA and included the dispositions for each source of uncertainty for this TSTF-505 application. Furthermore, as part of its audit of the LAR, the NRC staff reviewed the PRA analyses that support the uncertainty analysis for the LaSalle IEPRA, FPRA, and risk applications cited in the LAR. Upon review of the dispositions provided in LAR Tables E9-1, E9-2 and E9-3, it was unclear to the NRC staff how the licensee concluded there was no impact on the RICT calculations. In light of these observations, address the following:

a) LAR Enclosure 9, Table E9-3 identifies cable selection as a key source of FPRA modeling uncertainty because of conservatisms in the approach (i.e., lack of cable data).

The LAR discusses that an informed approach was used in developing the assumed

[cable] routing and that other modeling assumptions provide some offsetting effects; therefore, the licensee concludes that Unknown Location (UNL) modeling uncertainty has no impact on the RICT program calculations. It is not clear to the NRC staff how this assumption (i.e., cable selection) was concluded to have no impact on the RICT program calculations, given a sensitivity study demonstrated that its impact on risk is moderate. The NRC staff also notes that this conservatism in PRA modeling could have a non-conservative impact on the RICT calculations (e.g., potential to mask importance of another SSC).

Therefore, address the following:

i. Identify the systems or components that are assumed to be always failed in the FPRA or not included in the FPRA (due to lack of cable tracing, etc.) and provide sufficient justification to support the conclusions that the RICT calculations are not adversely impacted and no RMAs are required to address this item.

ii. As an alternative to part (i), above, propose a mechanism to ensure that a sensitivity study is performed for the RICT calculations for applicable SSCs

which account for the impact on the RICT. The proposed mechanism should also ensure that any additional risk from the assumption that the SSC is always failed is either accounted for in the RICT calculation or is compensated for by applying additional Risk Management Actions (RMAs) during the RICT (e.g., include as an implementation item in Attachment 5 of the LAR).

b) LAR Enclosure 9, Table E9-1, identifies vapor suppression capability following vessel rupture failure as a source of uncertainty. The LAR states that [a]lthough the RICT estimates change as a result of this sensitivity, the bounding sensitivity analysis utilizes the upper bound values, which is not a realistic assumption and use of this bounding assumption would result in overly conservative RICT estimates. It appears to the NRC staff that the results of the sensitivity study using upper bound values from NUREG/CR-6595 validate the concern that this source of uncertainty impacts the RICT calculations.

In light of the staff concerns, address the following:

i Provide the results of the cited sensitivity study that demonstrates which TS LCOs are impacted and what the change in the calculated RICT would be using the values from NUREG/CR-6595.

Ii Given the uncertainty indicated by recent MAAP runs cited in the LAR and the results of the sensitivity study on vapor suppression failure following vessel failure, provide sufficient justification to support the conclusion that the upper bound values used are not realistic and would result in overly-conservative RICT estimates. Include any RMAs identified that will be applied during the RICTs for the TS LCOs.

c) The LaSalle FPRA uncertainty analysis report reviewed by NRC staff during the audit provides the results of seven sensitivity studies. Case #2 regarding credit for the hardened containment vent system (HCVS) (refer to Section 4.2.2 of LS-PSA-021.12, Revision 3) is credited in the FPRA for both units. The sensitivity study results provided in the report demonstrate that the fire CDF is very sensitive to credit for the HCVS (i.e.,

the fire CDF increases by a factor of 3 if credit for the HCVS is removed). Neither this report or the LAR discuss the uncertainty associated with modeling the HCVS in the FPRA, nor do these documents discuss how this uncertainty will be treated in the RICT program. In light of the NRC staff concerns, address the following:

I Discuss the uncertainty associated with modelling the HCVS in the FPRA along with justification that the modelling of the HCVS is sufficient for this application given the impact that credit for HCVS has on the fire CDF.

Ii If in the response to part (i) above, the modelling of the HCVS in the FPRA cannot be justified for this application given the significant impact that HCVS modeling credit has on the fire CDF, propose a mechanism that addresses this source of modelling uncertainty in the RICT program (e.g., include as an implementation item in Attachment 5 of the LAR).

d) The LaSalle FPRA uncertainty analysis report reviewed by NRC staff during the audit provides discussion of sources of uncertainty documented in various FPRA notebooks.

The report identifies a source of fire scenario development uncertainty pertaining to visual identification of target sets and determination of whether they are judged to be

located within the zone of influence (ZOI) for fires modeled in that area (refer to Section 3.2.8 of LS-PSA-021.12, Revision 3, pages 3-15, Item #1). The report indicates that visual identification of targets and their location with respect to generic ZOIs is an uncertainty associated with generic ZOIs. Provide the following additional information:

i. Describe how the visual identification of fire targets located within the ZOIs of fires was performed. Include in the discussion sufficient justification to support the conclusion that the source of uncertainty has minimal impact on the RICT calculations.

e) LAR Enclosure 9, Table E9-1, identifies credit for survivability of the emergency core cooling system (ECCS) after containment venting as a key source of uncertainty. It is not clear to the NRC staff how this key source of uncertainty was concluded to have negligible impact on RICT calculations and no further treatment (i.e., RMAs) is needed.

The NRC staff acknowledges that increasing the conditional failure probability of the ECCS to 1.0 is bounding. Provide the following additional information:

i Discuss any applicable RMAs for the TS LCOs impacted by this uncertainty (e.g.,

TS LCO Conditions 3.3.5.1.C, 3.3.5.1.D, 3.5.1.A, and 3.5.1.C). Include justification for why the RMAs are sufficient to address this uncertainty, specifically discuss why additional RMAs or other measures are not needed to reduce the risk of the impacted TS LCOs.

ii If in response to part (i) above, it cannot be justified that the applicable RMAs are sufficient to address this uncertainty, then include additional RMAs or other measures that will be used to reduce the risk impact of the cited uncertainty for the applicable TS LCOs (e.g., include as an implementation item in Attachment 5 of the LAR).

APLA QUESTION 11 - Evaluating State-of-Knowledge Correlation Uncertainty Impact on the RICT Program As provided by the guidance in NEI 06-09, Revision 0-A, changes to CDF and LERF, as calculated by a PRA, that models the current plant operating configuration, are used to support the RICT program. The guidance in NEI 06-09, Revision 0-A, provides several quantitative risk management threshold values: the calculated RICT, the calculated instantaneous risk, and the cumulative risk increase. When a risk threshold value is exceeded, specific actions are required as summarized in Table 2.2 of NEI 06-09, Revision 0-A.

RG 1.174 clarifies that, because of the way the acceptance guidelines in RG 1.174 have been developed, the appropriate numerical measures to use when comparing the PRA results with the risk acceptance guidelines are mean values. The risk management threshold values for the RICT program have been developed based on RG 1.174 and, therefore, the most appropriate measures with which to make a comparison are also mean values.

Point estimates are the most commonly calculated and reported PRA results. Point estimates do not account for the state-of-knowledge correlation (SOKC) between nominally independent basic event probabilities, but they can be quickly calculated. Mean values reflect the SOKC and are always larger than point estimates, but require longer and more complex calculations.

NUREG-1855, Revision 1 provides guidance on evaluating how the uncertainty arising from the propagation of the uncertainty in parameter values (SOKC) of the PRA inputs impacts the

comparison of the PRA results with the guideline values. In light of these observations, address the following:

a) Provide a summary of how the SOKC investigation was performed for the base LaSalle PRA models used to support the RICT application.

b) Provide a summary of how the SOKC will be addressed for the RICT program (i.e.,

based upon the risk metrics to be considered), and explain how this process/approach is consistent with NUREG-1855, Revision 1.

APLA QUESTION 12: Key Principle 5 - Monitoring and Maintenance Rule Program Regulatory Basis:

Under 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired.

Section 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (i.e., the Maintenance Rule), requires licensees to monitor the performance or condition of SSCs against licensee established goals in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions.

The regulation in 10 CFR 50.65(a)(4) requires the assessment and management of the increase in risk that may result from a proposed maintenance activity.

Request:

Section 2.3 of LAR Attachment 1 states that the application of a RICT will be evaluated using the guidance provided in NEI 06-09, Revision 0-A, which was approved by the NRC on May 17, 2007 (ADAMS Accession No. ML071200238). The referenced NRC SE for NEI 06-09, Revision 0-A, states the following on page 20 of the SE associated with item 5, The impact of the proposed change should be monitored using performance measurement strategies:

The performance monitoring and feedback specified in the TR

[Topical Report], is sufficient to reasonably assure changes in risk due to the implementation of the RMTS are small, and are consistent with Section 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied.

Section 2.3 of LAR Attachment 1 also states:

In addition, the NEI 06-09-A, Revision 0 methodology satisfies the five key safety principles specified in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making:

Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176), relative to the risk impact due to the application of a RICT.

Key Principle 5, page 1.177-4, of the version of RG 1.177 dated August 1998, referenced in the LAR states the following:

The impact of the proposed change should be monitored using performance measurement strategies. The three-tiered implementation approach discussed in Regulatory Position 3.1 and Maintenance Rule control discussed in Regulatory Position 3.2 provide guidance in meeting this principle.

Regulatory Position 3.2, Maintenance Rule Control, of RG 1.177, dated August 1998, states the following:

To ensure that extension of a TS AOT [allowed outage time] or STI [surveillance test interval] does not degrade operational safety over time, the licensee should ensure, as part of its Maintenance Rule program (10 CFR 50.65), that when equipment does not meet its performance criteria, the evaluation required under the Maintenance Rule includes prior related TS changes in its scope.

If the licensee concludes that the performance or condition of TS equipment affected by a TS change does not meet established performance criteria, appropriate corrective action should be taken, in accordance with the Maintenance Rule. Such corrective action could include consideration of another TS change to shorten the revised AOT or STI, or imposition of a more restrictive administrative limit, if the licensee determines this is an important factor in reversing the negative trend.

Regulation 10 CFR 50.65 does not specifically identify the use of performance criteria to evaluate the performance of SSCs. The NRC endorsed the guidance contained in NUMARC 93-01, Revision F (ADAMS Accession No. ML18120A069), in RG 1.160, Revision 4 (ADAMS Accession No. ML18220B281). NUMARC 93-01, Section 9.0, Establishing Risk and Performance Criteria/Goal Setting and Monitoring, contains guidance for the establishment of performance criteria. NUMARC Section 11.3.7, Managing Risk, also addresses the use of SSC performance criteria with regard to 10 CFR 50.65(a)(4) risk assessments.

As described above, the references contained in the LaSalle LAR dated January 31, 2020, to adopt TSTF-505 cascade to credit conformance to Regulatory Position 3.2 in RG 1.177, dated August 1988. Therefore, the LaSalle TSTF-505 LAR credits the use of established performance criteria in the Maintenance Rule program as part of the justification basis for the LAR.

Question Confirm that the Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in the NRC-endorsed guidance in NUMARC 93-01.

Alternatively, describe the method used by LaSalle to establish performance criteria for SSC performance as described in Regulatory Position 3.2 in order to satisfy key principle 5 as referenced in RG 1.177, dated August 1998, and credited in the justification basis for the proposed LaSalle LAR dated January 31, 2020, to adopt TSTF-505.

APLC QUESTION 01 - Screening of Non-Seismic External Hazard Screening of Non-Seismic External Hazard 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 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.

External Flooding Section 5 of Enclosure 4 of Attachment 2 to the LAR discusses the evaluation of external flooding on the proposed RICTs and concludes that the risk from external flooding is considered negligible and can be screened from inclusion in the RICT program. The basis for screening the external flooding hazard includes the results documented in the licensee's focused evaluation based on the re-evaluated flood hazard for the site (Reference 25). The licensee states that flood mitigation was achieved by permanently installed passive flood protection (e.g., exterior doors and plant grade) that require no manual action for success. Enclosure 4 to the Attachment to the licensees LAR for adoption of 10 CFR 50.69 (ADAMS Accession No. ML20031E699) indicates that there are several flood doors integral to flood protection.

Section 3.3.5 of NEI 06-09-A states that In addition to the evaluation of external events for potential RICT impact, these events should be evaluated for insights which permit development and implementation of applicable risk management actions. The LAR does not describe any risk management actions to ensure that the flood protection features, which are integral to flood protection and important for screening of external flooding, continue to be available and functional during the proposed RICTs.

Discuss how the licensees proposed RICT program will ensure that assumptions related to the availability and the functionality of flood protection features (e.g., flood doors) 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 Winds Section 5 of Enclosure 4 of Attachment 2 to the LAR discusses the evaluation of high winds on the proposed RICTs and concludes that the risk from high winds is considered negligible and can be screened from inclusion in the RICT program. The discussion includes a description of procedures that would mitigate the potential impact of tornado-generated missiles on two non-conformances against the tornado missile licensing basis. The procedures include operators starting standby train(s) of ventilation if a single train were failed due to a tornado missile strike and opening doors and aligning temporary cooling with equipment (e.g., door stops, portable fans) staged in boxes within the plant.

Section 3.3.5 of NEI 06-09-A states that In addition to the evaluation of external events for potential RICT impact, these events should be evaluated for insights which permit development and implementation of applicable risk management actions. The LAR does not describe any risk management actions to ensure that the procedures and any necessary equipment identified

in the LAR for tornado-generated missiles on the two non-conformances will be able to provide mitigation during the proposed RICTs.

Discuss how the licensees proposed RICT program will ensure that assumptions related to the risk from high winds and tornado-generated missiles (e.g., effective procedures and availability of any necessary mitigating equipment, including the two SSCs currently identified as tornado-generated missile protection non-conformances), remain valid during RICTs such that the high winds risk (including risk from tornado-generated missiles) continue to have an insignificant impact on the configuration-specific risk.

Turbine Missiles The discussion for turbine missiles states the speed capability of these rotors is considerably higher than the maximum attainable speed of these turbine generator units. Consequently, the probability of missiles being generated is statistically insignificant. The turbine missile probability analysis evaluates the failure of turbine stop, control, and bypass valves and determines the inspections and frequency of those inspection so that the failure rate and probability of turbine missile damaging safety related equipment is below the threshold of 10-7.

It is unclear whether the turbine missile probability analysis is the basis for the screening or the speed capability of the rotors.

Clarify whether the basis for screening the turbine missile hazard is in the probability of turbine missile analysis. If not, provide justification for not using that analysis and selecting an alternate approach.

EICB QUESTION 01 The TSTF-505, Revision 2, position it that at least one redundant or diverse means (e.g., other automatic features or manual action) to accomplish the safety functions (e.g.,

reactor trip, safety injection, or containment isolation) remain available during the use of the RICT. of the LAR lists the FUNCTIONAL UNITS of the I&E systems; however, this list does not provide NRC staff adequate information to verify at least one redundant or diverse means will remain available to accomplish the intended safety functions of I&C TS with RICT.

3.3.1.1, 2.b (15.4.9) Control Rod Drop Accident 3.3.1.1, 4 (15.2.8/15.6.6) Feedwater Line Break Outside Containment 3.3.1.1, 4 (15.6.5) Loss of Coolant Accidents 3.3.1.1, 5 (15.2.4) Main Steam Isolation Valve - Closure 3.3.1.1, 5 (15.6.4) Steam System Piping Break Outside Containment 3.3.1.1, 5 (15.1.3) Pressure Regulator Failure 3.3.1.1, 6 (15.6.5) Loss of Coolant Accidents 3.3.1.1, 7.a (None) 3.3.1.1, 7.b (None) 3.3.1.1, 8 (15.1.2A) Feedwater Controller Failure - Maximum Demand 3.3.1.1, 8 (15.1.3) Pressure Regulator Failure - open 3.3.1.1, 8 (15.2.3/15.2.3A) Turbine Trip 3.3.1.1, 8 (15.2.5) Loss of Condenser Vacuum 3.3.1.1, 8 (15.2.6) Loss of A-C Power

3.3.1.1, 8 (15.3.1) Recirculation Pump Trip 3.3.1.1, 8 (15.3.2) Recirculation Flow Control Failure - Decreasing Flow 3.3.1.1, 8 (15.3.3) Recirculation Pump Seizure 3.3.1.1, 8 (15.2.9) Residual Heat Removal Shutdown Cooling 3.3.1.1, 9 (15.2.2A) Generator Load Rejection 3.3.1.1, 9 (15.3.4) Recirculation Pump Shaft Break The column for Diverse Instrumentation, some of the initiation signals are not discussed in UFSAR. The evaluation of diverse means should identify the conditions that the FUNCTIONAL UNIT responds to, and for each condition, other means (e.g., diversity, redundancy, or operator actions) that can be used. Alternatively, provide additional information to demonstrate that defense in depth is maintained during the extended completion times for each function. This information is needed to demonstrate compliance with 10 CFR 50.36(c), and consistent with the implementing guidance in RG 1.174 and the TSTF-505 Revision 2.

A number of diverse means are identified as manual actuation. Please confirm that these manual actuations identified in Attachment 7 are modeled in PRA, This applies to the entire list of TS tables identified in Attachment 7.

EEOB QUESTION OFFSITE AND ONSITE POWER SYSTEMS

Background

The LAR proposes changes to LCO associated with TS 3.8.1 AC Sources-Operating for onsite and offsite power systems. The licensing basis of LaSalle is a loss-of coolant accident (LOCA) in one unit while the remaining unit is being shut down without loss of coolant, as well as a condition in which both units are concurrently being shut down without LOCAs, with or without offsite power.

The NRC staff notes that in the event of a dual unit trip, turbine stop valve closure for each turbine may initiate an accident or safety system actuation signal in both units.

Diesel generator (DG) 0 (Division 1) is shared between Units 1 and 2. One DG is permanently assigned to each of the three engineered safety features electrical system 4160-volt buses for each unit. The automatic transfer switch for DG 0 seeks Unit 1, Division 1 but an accident signal prioritizes the alignment of DG 0.

The PRA and design success criteria for LCO 3.8.1.A, in Table E1-1 of Enclosure 1 states One offsite source and LCO 3.8.1.B in Table E1-1 states one diesel per division LCOs 3.8.1.C through E refer to these LCOs.

Discussion Please provide a discussion on:

a) The applicability of TS LCOs associated with the DGs for dual unit operation.

b) Modelling of the systems and the success paths used to calculate the RICTs for the associated LCOs with specific reference to shared DG 0 and opposite Unit Division 2 DG. Please include a discussion on shared systems associated with Division 2 DG.

c) Please include a discussion on transformer capabilities and any cross connections (if applicable include timelines) that were credited for controlled shutdown of dual units using offsite power.

d) Action Statement 3.8.1.D is associated with potential loss of two required offsite power circuits inoperable. For Action 3.8.1.D, a 30-day RICT is estimated in Table E1-2 of Enclosure 1. The NRC staff notes that during normal operation, core standby cooling system-equipment cooling water system (CSCS-ECWS) circulates lake water from the ultimate heat sink (UHS) for cooling of various heat exchangers including diesel-generator coolers. According to the LaSalle UFSAR, the CSCS-ECWS for each unit is designed as three independent subsystems, one of which is shared between units. The CSCS-ECW subsystems take a suction from the service water tunnel and the service water tunnel is kept full by six inlet lines which connect to the circulating water pump forebays. If the design of UHS uses non-essential service water systems to maintain UHS inventory to ensure 30-day mission time of UHS assumed in accident analysis, then provide a discussion on modeling and any impact on the UHS and associated cooling water systems during entry into these LCOs and heat removal capability of the systems.

e) The two required qualified circuits for each unit share common system auxiliary transformers and a common switchyard. Hence, there is a potential for both units to simultaneously be in the LCO associated with Action Statements 3.8.1.D. Please provide a discussion on the RICT modelling for dual unit operation and any impact on shared systems during dual unit entry into LCO 3.8.1.D. Please include a discussion on PRA success criteria that may rely on an operable system in the opposite unit.

f) DG 0 (Division 1) is critical for the safety functions of both LaSalle units. Please discuss its and the associated breaker closing circuit/logic reliability for the last 5 years based on operation and maintenance?

EEOB QUESTION INSTRUMENTATION.

Background

Instrumentation is generally not modeled in detail in PRAs and is often modeled as a single, generic basic event, representing all trains or channels. Several instrumentations related LCOs are associated with One or more channels inoperable. In some cases, a Note stating Not applicable when a loss of function occurs is proposed. In some cases, there is no Note;. there are also LCOs where the note is related to only one of the ACTION Statements. As an example, LCO 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation function listed below shall be OPERABLE. CONDITION E and F have COMPLETION TIMES (CTs) of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and 8 days. The NOTE for loss of function appears to be applicable to the CT of 8 days only.

The LAR proposes changes to LCO 3.3.8.1, Loss of Power (LOP) instrumentation. The NRC staff noted the following design basis information on loss of power (LOP) relays for the LaSalle Design:

For Divisions 1 and 2, each loss of voltage and degraded voltage function is monitored by two instruments per bus whose output trip contacts are arranged in a two-out-of-two logic configuration per bus. The loss of voltage signal is generated when a loss of voltage occurs for a specific time interval.

For Division 3, the degraded voltage function logic is the same as for Divisions 1 and 2, but the Division 3 loss of voltage function logic is different. The Division 3 DG will auto-start if either one of the two bus undervoltage relays (with a time delay) actuates and the DG output breaker will automatically close with the same undervoltage permissive provided that the Division 3 bus main feeder breaker is open and the DG speed and voltage permissive are met.

For Divisions 1 and 2, LCO 3.3.8.1 requires two channels per bus of the loss of voltage Function and two channels per bus of the degraded voltage Function to be OPERABLE.

The Division 3, degraded voltage relay scheme has a similar logic. The NRC staff notes that the logic for proper operation of the LOP instrumentation requires 2 out of 2 channels detecting and actuating the associated relays. Failure of one channel will result in failure to start the associated DG and failure to separate the safety related busses from an inoperable offsite power source rendering the unit with loss of both, onsite and offsite power systems, on the applicable train.

Discussion a) Describe how, One or more functions with one channel or one or more buses inoperable precludes loss of function as considered in the RICT analysis for the proposed changes to instrumentation related LCOs.

b) For each condition in TS 3.3.8.1 proposed in the scope of the RICT program (including time to place channel in trip condition), describe how it is modelled in the PRA. Specifically, for LCOs 3.3.8.1A associated with LOP channels, please discuss the consequences of failure in one channel and potential delay to default the channel into a trip condition.

c) For Division 3, the loss voltage logic requires only one channel to perform the intended function. Describe how this logic in LCO 3.3.8.1 is applied to Division 3 relays and the consequences of delay to default the channel into trip mode.

d) Describe the intent of two out of two logic per division when used as design success criteria. (Reference Table E1-1: In Scope TS/LCO conditions to corresponding PRA functions) e) Since the offsite power and the onsite DG associated with an inoperable channel may not be available, describe the adequacy of assuming a surrogate relay is chosen that only fails the DG start mode for an inoperable undervoltage relay.

(Reference Table E1-1: In Scope TS/LCO conditions to corresponding PRA functions) f) Discuss how one or more channels is applied to channels in redundant or diverse systems measuring similar parameters in different divisions - e.g. Loss of voltage function in Division 1 and degraded voltage function in Division 2 of the same unit or the other unit.

EEOB QUESTION AC and DC SYSTEMS

Background

The NRC staff notes that control power for each DG is supplied from the 125-VDC battery within its associated division. The 125-VDC control power for DG "0" is supplied from either Unit 1 Division 1 or Unit 2 Division 1 as determined by the position of an automatic transfer switch located in the diesel generator "0" control panel. The automatic transfer switch seeks Unit 1, Division 1.

LCO 3.8.4.B is associated with Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Condition A and has a RICT estimate of 3.1 days. LCO 3.8.7.A is associated with one or both Division 1 and 2 AC electrical power distribution subsystem inoperable and has a RICT estimate of 0.0 days.

Discussion Please provide a discussion on how the 125 VDC system is modelled in PRA with specific discussion on the DC control power for operability of the associated DGs and power distribution system considered in LCO 3.8.7 A.

ML20160A164 *by e-mail concurrence OFFICE DORL/LPL3/PM DORL/LPL3/LA* DRA/APLA/BC* Dex/EICBB/BC* DRA/APLC/BC*

NAME BVaidya SRohrer RPascarelli MWaters SRosenberg DATE 06/10/2020 06/10/2020 06/05/2020 06/10/2020 06/05/2020 OFFICE DEX/EEOB/BC* DORL/LPL1/BC* DORL/LPL3/PM NAME BTitus NSalgado BVaidya DATE 06/09/2020 06/11/2020 06/11/2020