ML22070B130

Summary of Regulatory Audit in Support of Risk-Informed Completion Times in Technical Specifications License Amendment Request

ML22070B130
Person / Time
Site:	Susquehanna
Issue date:	03/31/2022
From:	Audrey Klett Plant Licensing Branch 1
To:	Cimorelli K Susquehanna
Klett A
References
EPID L-2021-LLA-0062
Download: ML22070B130 (50)

Text

March 31, 2022 Mr. Kevin Cimorelli Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard NUCSB3 Berwick, PA 18603-0467

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 -

SUMMARY

OF REGULATORY AUDIT IN SUPPORT OF RISK-INFORMED COMPLETION TIMES IN TECHNICAL SPECIFICATIONS LICENSE AMENDMENT REQUEST (EPID L-2021-LLA-0062)

Dear Mr. Cimorelli:

By letter dated April 8, 2021,1 Susquehanna Nuclear, LLC (the licensee) submitted a license amendment request (LAR) for Susquehanna Steam Electric Station, Units 1 and 2.2 The licensee proposed changes to various technical specifications (TS) to permit the use of risk-informed completion times (RICTs) in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times. 3 The licensee also proposed variations from TSTF-505, Revision 2 and TS changes not associated with TSTF-505, Revision 2.

The U.S. Nuclear Regulatory Commission (NRC) staff conducted a virtual audit to support its review of the LAR. The NRC staff audited various licensee documents and interviewed licensee staff and representatives. The NRC staff issued its audit plan on June 15, 2021.4 of this audit summary lists the individuals that took part in or attended the audit. lists the NRC staffs audit requests and questions. Enclosure 3 contains background information supporting some of the audit requests and questions. Enclosure 4 lists the documents the NRC audited.

1 PLA 7897, Agencywide Documents Access and Management system (ADAMS) Accession No. ML21098A206 2 Renewed Facility Operating License Nos. NPF-14 and NPF-22, respectively 3 ADAMS Accession No. ML18183A493, dated July 2, 2018 4 ADAMS Accession No. ML21153A137

The NRC staff conducted the audit using virtual meetings and an internet-based portal provided by the licensee. Using the licensees portal, the NRC staff reviewed documents (e.g., calculations and reports) related to the LAR but not available on the Susquehanna dockets. During the audit, the staff also met virtually with the licensee approximately every other week. The staff used these meetings to confirm its understanding of the LAR, discuss the documents in the portal, and determine whether the NRC staff identified any information that needs to be submitted on the docket to complete the NRC staffs safety evaluation.

After the audit discussions, the licensee supplemented its LAR on March 8, 2022.5 The NRC staff is reviewing the licensees supplement to decide if the NRC needs any additional information to complete its review of the licensees request.

In lieu of an exit meeting, the NRCs licensing project manager informed licensee staff by telephone on January 31, 2022, that the NRC staff had completed its audit.

If you have any questions, please contact me at (301) 415-0489 or by e-mail to Audrey.Klett@nrc.gov.

Sincerely,

/RA/

Audrey L. Klett, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosures:

1. List of Audit Participants

2. List of Audit Questions and Requests

3. Background Document

4. List of Audited Documents cc: Listserv 5 PLA 7984, ADAMS Accession No. ML22067A171

List of Audit Participants U.S. Nuclear Regulatory Commission and Contract Staff Licensee and Contract Staff Ballard, Brent (NRR1/DORL2/LPL13)

Andersen, Vince (Jensen Hughes)

Beck, Tyler (NRR/DORL/LPL2-14)

Boyer, Jacob (Susquehanna5)

Coles, Garill (PNNL6)

Brown, Katie (Susquehanna)

Correll, Brian (R-IV/DRS/IPAT7)

Hartzell, Jason (Susquehanna)

Grenier, Bernie (NRR/DRA8/APLB9)

Hoppe, David (Susquehanna)

Grover, Ravi (NRR/DSS10/STSB11)

Hufnal, Jared (Jensen Hughes)

Highley, Christopher (R-I/DORS/PB412)

Jurek, Shane (Susquehanna)

Hilsmeier, Todd (NRR/DRA/APLA13)

Kolonauski, Lynn (Jensen Hughes)

Iqbal, Naeem (NRR/DRA/APLB)

Krick, Melissa (Susquehanna)

Khan, Nadim (NRR/DEX14/EEEB15)

Lee, Larry (Jensen Hughes)

Klett, Audrey (NRR/DORL/LPL1)

McCabe, Kip (Susquehanna)

Lentchner, Gabe (NRR/DSS/STSB)

Miller, Drew (Jensen Hughes)

Li, Ming (NRR/DEX/EICB16)

Radford, Todd (Jensen Hughes)

Miller, Ed (NRR/DORL/LPL2-1)

Shappaugh, Jeff (Jensen Hughes)

Moulton, Charles (NRR/DRA/APLB)

Shanley, Leo (Jensen Hughes)

Nguyen, Khoi (NRR/DEX/EEEB)

Shaw, Erick (Jensen Hughes)

Park, Sunwoo (NRR/DRA/APLC17)

Sternowski, Nicholas (Jensen Hughes)

Pascarelli, Robert (NRR/DRA/APLA)

Sweeney, Timothy (Susquehanna)

Patterson, Malcolm (NRR/DRA/APLA)

Vanover, Donald (Jensen Hughes)

Rossi, Matthew (R-I/DORS/PB4)

Vazquies, Ronald (Susquehanna)

Scott, Christian (NRR/DSS/STSB)

Wishart, Mark (Jensen Hughes)

Tetter, Keith (NRR/DRA/APLC)

Yenchak, Joseph (Susquehanna)

Wagage, Hanry (NRR/DSS/SCPB18)

West, Khadijah (NRR/DSS/STSB)

Other Zhao, Jack (NRR/DEX/EICB)

Shields, Matthew (State of Pennsylvania) 1 Office of Nuclear Reactor Regulation 2 Division of Operating Reactor Licensing 3 Plant Licensing Branch 1 4 Plant Licensing Branch 2-1 5 Susquehanna Nuclear, LLC 6 Pacific Northwest National Laboratory (Contractor) 7 Region IV Office/Division of Reactor Safety/Inspection, Program and Assessment Team 8 Division of Risk Assessment 9 Probabilistic Risk Assessment (PRA) Licensing Branch B 10 Division of Safety Systems 11 Technical Specifications Branch 12 Region I Office/ Division of Operating Reactor Safety/Projects Branch 4 13 PRA Licensing Branch A 14 Division of Engineering and External Hazards 15 Electrical Engineering Branch 16 Instrumentation and Controls Branch 17 PRA Licensing Branch C 18 Containment and Plant Systems Branch

List of Audit Questions and Requests The U.S. Nuclear Regulatory Commission (NRC) staff requested the following information during its audit to support its review of the license amendment request (LAR) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21098A206) for the Susquehanna Steam Electric Station (Susquehanna, or SSES), Units 1 and 2.

No.

Requested Documents (D) and Audit Questions (Q)

D-1 Near-term (i.e., late June or early July 2021): Presentation of an overview of the risk-informed completion time (RICT) program procedures, including those for determining risk management actions (RMAs), and a demonstration of the process for calculating RICTs (including use of the configuration risk management program tool)

D-2 Results of the fire probabilistic risk analysis (PRA)

D-3 Reports of full-scope and focused-scope peer reviews (and facts and observations (F&Os) closure reviews referred by date in the LAR, Enclosure 2, Sections 3, 4, and 5) and any self-assessment performed for the internal events, internal flooding, and fire PRAs cited in the LAR D-4 Closure reports on F&Os from these assessments D-5 For the internal events, internal flooding, and fire PRAs, plant-specific documentation (e.g., uncertainty notebooks) related to:

a. The review of the PRA model assumptions and sources of uncertainty

b. Identification of key assumptions and sources of uncertainty for the application D-6 PRA notebooks for the modeling of FLEX equipment and FLEX human error probabilities credited in the PRA D-7 Documentation supporting the example RICT calculations presented in LAR,, Table E1-2 D-8 Any draft or final RICT program procedures (e.g., for RMAs, PRA functionality determination, and recording limiting conditions for operation)

D-9 Plant and PRA configuration control procedures D-10 Documentation supporting the development of the real-time risk tool and benchmarking it against the PRA D-11 Relevant design documentation, e.g., single line diagrams of the electrical power distribution systems and piping & instrumentation diagrams D-12 Design details of systems shared or cross-tied between Unit 1 and Unit 2, including electrical and mechanical systems D-13 Documentation of how shared or cross-tied systems are modeled in the PRA D-14 The following documents referenced in the Susquehanna LAR, if not already identified:

a. Susquehanna Calculation EC-RISK-0043, Technical Adequacy of the SSES PRA Models - Probabilistic Risk Assessment, Revision 1, dated December 7, 2020

b. Susquehanna Calculation EC-RISK-0040, Summary and Quantification of Model SSES19R0I1 - Full Power Internal Events Probabilistic Analysis (FPIE), Revision 0, dated August 3, 2020

c. Susquehanna Calculation EC-RISK-0056, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications, Revision 0, dated February 8, 2021

d. Appendix B of the roadmap document (NQPA-B-NA-009)

No.

Requested Documents (D) and Audit Questions (Q)

e. Results Notebook (PA-BNA-232)

f. Parametric uncertainty notebooks for internal events, internal flooding, and fire PRAs D-15 All primary Susquehanna fire PRA notebooks D-16 The following calculations:

a. Susquehanna Calculation EC-RISK-0048, Fire PRA Interim Model OCT17R2F1 Quantification and Sensitivity - Fire Probabilistic Risk Analysis (FPRA), Revision 0, dated September 22, 2020.

b. Susquehanna Calculation EC-RISK-0046, Seismic CDF [Core Damage Frequency]

and LERF [Large Early Release Frequency] Estimate for TSTF [Technical Specifications Task Force]-505 (RICT) Program - Seismic Margin Analysis for Risk Informed Applications, dated August 28, 2020.

c. Susquehanna Calculation EC-RISK-0045, High Confidence Low Probability Failure (HCLPF) Value for the Seismic Penalty Calculation - Seismic Margin Analysis for Risk Informed Applications, dated August 28, 2020.

d. Susquehanna Calculation EC-RISK-0047, External Hazards Assessment - SSES External Hazards Assessment for Risk Informed Applications, Revision 0, dated August 28, 2020.

e. Susquehanna Calculation EC-093-1023, Turbine Missile Probability Analysis for Susquehanna Unit 1 & 2, Revision 3, dated August 29, 2014.

D-17 Other documentation that the licensee determines to be responsive to the staff's information requests D-18 (now Q-1)

D-19 Load list for each safety-related bus D-20 Plant procedures related to the RMA for the electrical power systems, if available D-21 Excel Workbook, Susquehanna SCDF-SLERF [Seismic CDF-Seismic LERF]

Estimate_Rev0.xls, dated December 2019 D-22 Excel Workbook, PA-B-NA-502_R2.xls re. calculations of SCDF and SLERF Q-1 Additional justification for adding following isolation in the completion times for Actions A.2, C.2, and D.2 in TS 3.6.1.3.

Q-2 Refer to Background Document Open Internal Events PRA Facts and Observations EITHER:

Confirm that F&O 1-18 has been resolved by showing that [the licensee] has (1) reviewed and updated the list of internal events PRA modeling assumptions and sources of uncertainty based on the disposition of the 2020 F&O closure review team, (2) provided dispositions that are specific to this application, and (3) addressed, in accordance with Nuclear Energy Institute (NEI) 06-09, Revision 0-A (or NEI 06-09-A, ADAMS Accession No. ML122860402), any assumptions or sources of uncertainty determined to be potentially key to this application (e.g., performed a sensitivity study demonstrating that they have an inconsequential impact on the RICT calculations, or identified programmatic changes to compensate for this modeling uncertainty). Also, confirm whether Report EC-RISK-0056, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications dated February 8, 2021, which predates the LAR dated April 8, 2021, represents the review update of the internal events PRA uncertainty analysis.

No.

Requested Documents (D) and Audit Questions (Q)

OR:

If the licensee cannot confirm that F&O 1-18 has been resolved per the above, then explain how the licensee will ensure (e.g., via a license condition or an implementation requirement), prior to implementation of the Risk-Managed Technical Specifications (RMTS) program, that it will: (1) review and update the list of internal events PRA modeling assumptions and sources of uncertainty, (2) update the associated dispositions that are specific to this application, and (3) fully address any impacts on the RMTS program (including those determined to be key for this application) in accordance with NEI 06-09, Revision 0-A.

Q-3 Refer to Background Document Dispositions of PRA Model Assumptions and Sources of Uncertainty

a. Describe how the plant configurations (i.e., technical specification (TS) limiting condition for operability (LCO) conditions) were chosen in the sensitivity studies to assess the impact of potential key assumptions and sources of uncertainty. Also include in this discussion the bases for selecting these plant configurations.

b. Justify how the process described in part (a), above, is sufficient to conclude that, based on the sensitivity study RICT estimates, the impact of the associated modeling uncertainty on the RICT calculations is inconsequential.

Q-4 Refer to Background Document Dispositions of PRA Model Assumptions and Sources of Uncertainty EITHER:

Report the results of a sensitivity study in which the increase in the operator failure probability in the sensitivity case is set low enough that it is not unrealistic but high enough that it tests the modeling uncertainty to demonstrate that this modeling uncertainty is not key for the RMTS program and has an inconsequential impact on the RICT calculations. Also, describe this sensitivity study and justify the appropriateness of the selected operator failure probability used in the sensitivity case.

Provide the bases for the chosen plant configurations (i.e., TS LCO conditions) in this sensitivity study.

OR:

Describe the programmatic changes to compensate for this modeling uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). This discussion should also identify the TS LCO conditions in scope of RMTS for which the RICT calculations are impacted by this uncertainty and discuss how the RICTs are impacted (e.g., describe and provide the results of applicable sensitivity studies). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) provide RMA examples that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

OR:

Provide a detailed justification (e.g., propose an implementation item to update the PRA to address this modeling uncertainty and discuss how this update addresses this uncertainty; describe and provide the results of a different type of sensitivity study) that this modeling uncertainty does not need to be addressed in the RMTS program as required by Section 2.3.4 of NEI 06-09, Revision 0-A.

No.

Requested Documents (D) and Audit Questions (Q)

Q-5 Refer to Background Document Credit for FLEX Equipment and Actions Describe the FLEX strategies credited for the internal events (including internal flooding) and fire PRA models.

Q-6 Refer to Background Document Credit for FLEX Equipment and Actions Identify the FLEX equipment credited and whether that equipment is portable or permanently installed, and:

a. Discuss whether the credited FLEX equipment (regardless of whether it is portable or permanently installed) is similar to other plant equipment credited in the PRA (e.g., systems, structures, and components (SSCs) with sufficient plant-specific or generic industry data).

b. For credited FLEX equipment that is not similar to other plant equipment credited in the PRA:

i. Discuss the data and failure rates used to support its modeling and provide the rationale for using the chosen data and any conservative adjustments that were made to the generic reliability values for similar equipment. Discuss whether the uncertainties associated with the parameter values are in accordance with the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)

PRA Standard, as endorsed by NRC Regulatory Guide (RG) 1.200, Revision 2 (ADAMS Accession No. ML090410014).

ii. Describe the sensitivity study performed to assess the impact of uncertainty associated with equipment failure probabilities on calculated RICTs and present the results of that study. Justify how the increase in equipment failure probabilities used in the sensitivity case constitutes bounding realistic estimates. Also, discuss the bases for the chosen TS LCO conditions in the sensitivity study. Because the 30-day RICT back-stop condition could mask the impact of this uncertainty in the sensitivity study, discuss whether the RICTs for plant configurations involving more than one LCO entry (e.g., where the calculated RICTs are less than the 30-day backstop) are significantly impacted by this uncertainty.

iii. Discuss whether the uncertainty associated with equipment failure probabilities is a key source of uncertainty for the RMTS program. If this uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed in the RMTS program (e.g., programmatic changes such as identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) for those TS LCOs in LAR Enclosure 12, Risk Management Action Examples, that are significantly impacted by this uncertainty, provide updated RMAs that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

Q-7 Refer to Background Document Credit for FLEX Equipment and Actions

a. Identify the FLEX operator actions credited in the PRA and discuss whether any of these operator actions contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06 (ADAMS Accession No. ML16286A297).

No.

Requested Documents (D) and Audit Questions (Q)

b. For credited operator actions related to FLEX equipment that contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06:

i. Describe the sensitivity study performed to assess the impact of uncertainty associated with FLEX human error probabilities (HEPs, both the FLEX independent and dependent HEPs) on calculated RICTs and present the results of that study.

Justify how the increase in the FLEX HEPs used in the sensitivity case constitutes bounding realistic estimates. Also, discuss the bases for the chosen TS LCO conditions in the sensitivity study. Because the 30-day RICT back-stop condition could mask the impact of this uncertainty in the sensitivity study, discuss whether the RICTs for plant configurations involving more than one LCO entry (e.g., where the calculated RICTs are less than the 30-day backstop) are significantly impacted by this uncertainty.

ii. Discuss whether the uncertainty associated with FLEX HEPs is a key source of uncertainty for the RMTS program. If this uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed in the RMTS program (e.g., programmatic changes such as identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) for those TS LCOs in LAR Enclosure 12, Risk Management Action Examples, that are significantly impacted by this uncertainty, provide updated RMAs that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

Q-8 Refer to Background Document Credit for FLEX Equipment and Actions Given the challenges of modeling FLEX mitigation strategies, explain whether the review of the FLEX modeling was included in the last peer review of the PRA models.

If it was not, then justify how the model changes associated with incorporating FLEX mitigating strategies does not constitute a PRA upgrade as defined in Section 1-2 of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2.

Q-9 Refer to Background Document Impact of State-of-Knowledge Correlation (SOKC) Uncertainty the RICT Calculations:

a. Provide a summary of how the SOKC investigation was performed for the base Susquehanna PRA models used to support the RMTS application. Provide and discuss the results of this SOKC investigation and whether the SOKC uncertainty has a significant impact on the RICT calculations.

b. Provide and discuss the results of a comparison study between the RICT values calculated using point estimate risk versus mean risk for various LCO conditions in scope of RMTS. The LCO conditions selected for this comparison study should have a point estimate RICT less than 30 days (i.e., the 30-day backstop does not mask the comparison results) and are considered most likely to be impacted by the SOKC uncertainty (i.e., point estimate RICT versus mean RICT). Provide the bases for the chosen LCO conditions in this comparison study. Also, provide the intermediate risk results from these RICT calculations (e.g., the CDFs and LERFs for the baseline case using point estimates and sensitivity case using mean values from the internal events, including internal flooding, and fire PRAs).

No.

Requested Documents (D) and Audit Questions (Q)

c. Based on the results above, provide a summary of how the SOKC will be addressed for RICT calculations during RMTS implementation (i.e., based upon the risk metrics to be considered), and explain how this process or approach is consistent with NUREG-1855, Revision 1 (ADAMS Accession No. ML17062A466).

Q-10 Refer to Background Document Total Risk and Accounting for the SOKC EITHER:

Demonstrate that the total risk for Susquehanna Units 1 and 2 is in conformance with RG 1.174, Revision 3 (ADAMS Accession No. ML17317A256) risk acceptance guidelines (i.e., CDF < 1E-4 and LERF < 1E-5 per year) after the total mean internal events (including internal flooding) and fire CDF and LERF values are calculated to account for the SOKC and for potential changes in risk due to any updates to PRA models performed in response to NRC staff requests. Identify the fire PRA parameters that are assumed to be correlated in the parametric uncertainty analysis of fire events (e.g., fire ignition frequencies, non-suppression probabilities, severity factors, spurious probabilities, fire human error probabilities), as well as the sources used for the associated uncertainty distributions (e.g., NUREG-2169 (ADAMS Accession No. ML15016A069), NUREG/CR-1278 (ADAMS Accession No. ML071210299), NUREG/CR-7150 (https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7150/index.html), and EPRI Human Reliability Analysis (HRA) Calculator uncertainty distributions).

OR:

Explain how the licensee will ensure (e.g., via a license condition or implementation requirement) that, prior to implementation of the RMTS program: (1) the total mean internal events (including internal flooding) and fire CDF and LERF will be calculated to account for the SOKC and updates to PRA models performed in response to NRC staff requests; and (2) the updated total risk (including seismic risk) values are still in conformance with the RG 1.174 risk acceptance guidelines (i.e., CDF < 1E-4 and LERF < 1E-5 per year).

Q-11 Refer to Background Document Consideration of Shared Systems in RICT Calculations Explain whether shared systems are credited in the internal events (including internal flooding) and fire PRA models for both units that support the RICT calculations and, if so, then (1) identify those systems, and (2) either explain how the shared systems are modeled for each unit in a dual unit event demonstrating that shared systems are not over-credited in the PRA models, or if the PRA models do not address the impact of events that can create a concurrent demand for the system shared by both units, then justify that this exclusion has an inconsequential impact the RICT calculations.

Q-12 Refer to Background Document Impact of Seasonal Variations Explain how the impact of seasonal variations on the PRA modeling will be evaluated as needed during a RICT evolution and justify that this approach is consistent with the guidance in NEI 06-09-A and its associated NRC safety evaluation.

Q-13 Refer to Background Document In-Scope LCOs and Corresponding PRA Modeling

a. LAR Table E1-1 states for TS LCO 3.5.1, ECCS [Emergency Core Cooling System] - Operating, Condition D, HPCI [High Pressure Coolant Injection] System Inoperable, that both the design basis and PRA success criterion is one of one train (i.e., one HPCI pump). It appears that LCO 3.5.1, Condition D defeats the design No.

Requested Documents (D) and Audit Questions (Q) basis success criterion and, therefore, represents a TS loss-of-function. TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493) does not authorize determination of a RICT when the condition represents a TS loss-of-function. Therefore, explain why LCO 3.5.1, Condition D does or does not represent a TS loss-of-function. Include clarification of the design basis success criteria for the HPCI system.

09/15 Update:

b. The equipment in the Design Success Criteria column of Table E1 cannot be the equipment in the condition statement that is inoperable. The staff requests a correction to Table E1.

c. If the design basis and PRA success criteria are not consistent, then the staff requests the licensee to explain the basis for the differences and the justification for the PRA success criteria.

Q-14 Refer to Background Document In-Scope LCOs and Corresponding PRA Modeling LAR Table E1-1 states for TS LCO 3.5.1, ECCS - Operating, Condition B, One LPCI [low pressure coolant injection] pump in one or both LPCI subsystems inoperable, that the PRA success criteria are generally consistent with the design basis. However, the table also shows that for a loss-of-coolant accident (LOCA) in the bottom head, the PRA success criterion is one RHR pump in each division.

This appears to be more stringent than the design basis success criterion (i.e., one-of-four LCPI pumps) and associated with a specific accident sequence.

Therefore, explain this apparent inconsistency and confirm whether the more stringent PRA success criterion is for one possible low likelihood event.

Q-15 Refer to Background Document In-Scope LCOs and Corresponding PRA Modeling LAR Table E1-1 states for TS LCO 3.7.2, Emergency Service Water (ESW) System, Conditions A, B, and C, that the success criteria are consistent with the design basis.

However, the table indicates that the design basis success criterion is one ESW pump in each loop, and the PRA success criterion is one of two subsystems. The design basis and PRA success criteria do not appear to be equivalent. Therefore, explain how the design basis and PRA success criteria are consistent based on the wording in Table E1-1. If the licensee cannot confirm that the design basis and PRA success criteria are consistent, then explain the basis for the difference and the justification for the PRA success criteria.

10/5 Update:

Provide updated Table E1 to clarify terms (division, loop, subsystem, etc.).

Q-16 Refer to Background Document PRA Modeling of Instrumentation and Controls (I&C)

Explain how I&C equipment that is applicable or impacts the RICT calculations is modeled or 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.,

whether all channels of an actuation circuit are modeled), (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.

No.

Requested Documents (D) and Audit Questions (Q)

Q-17 Refer to Background Document Uncertainty Associated with Digital Instrumentation and Control Modeling EITHER:

Describe and provide the results of a sensitivity study performed for each digital system modeled in the PRA demonstrating that the uncertainty associated with PRA modeling the digital I&C systems has inconsequential impact on the RICT calculations.

OR:

Identify the LCOs impacted by digital I&C system modeling and 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 requires additional RMAs.

Q-18 Refer to Background Document PRA Update Process 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.

Q-19 Refer to Background Document Performance Monitoring EITHER:

Confirm that the Susquehanna Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in the NRC-endorsed guidance in NUMARC 93-01 (ADAMS Accession No. ML18120A069).

OR:

Describe the approach or method used by Susquehanna for SSC performance monitoring, as described in Regulatory Position C.3.2 of RG 1.177, Revision 2 (ADAMS Accession No. ML20164A035), for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative or quantitative), along with the appropriate risk metrics, and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC safety evaluation for NEI 06-09-A.

Q-20 Refer to Background Document Open FPRA F&Os EITHER:

Justify that the incomplete fire PRA modeling associated with multiple spurious operation (MSO) scenario 2aj has an inconsequential impact on the calculated RICTs.

OR:

If the licensee cannot justify that the incorporation of MSO scenario 2aj has an inconsequential impact on the calculated RICTs, then describe how the licensee will ensure that MSO scenario 2aj is incorporated into the fire PRA model prior to implementation of the RICT program.

Q-21 Refer to Background Document Open FPRA F&Os EITHER:

Discuss the PRA components that are supported by the unlocated cable, and justify (preferably using the sensitivity recommended by the peer-reviewers) that the treatment of the unlocated cable has an inconsequential impact on the calculated RICTs.

No.

Requested Documents (D) and Audit Questions (Q)

OR:

If the licensee cannot justify that the treatment of the unlocated cable has an inconsequential impact on the calculated RICTs, then explain how the licensee will either ensure that the cable is located and properly modeled in the FPRA prior to implementation of the RICT program or identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09-A.

Q-22 Refer to Background Document Open FPRA F&Os EITHER:

Identify the active partitions that have been credited in the FPRA and provide justification that they will perform reliably in the accident scenarios for which they are credited. Include discussion of systems that rely on supporting systems, such as alternating current (AC) power or a water supply, to perform their functions. If barriers such as dampers and doors are considered to be active partitions, then identify the mechanism that changes the position of the door or damper (e.g., fusible link plus gravity, operator action) to confirm that the mechanism does not rely on an active support system.

OR:

Explain how removal of the credit taken for active partitions has an inconsequential impact on the RICT calculations.

Q-23 Refer to Background Document Update of FPRA with Internal Event F&O Resolutions EITHER:

Confirm that all internal events modeling updates performed to resolve internal event F&Os that could impact fire risk were incorporated into the FPRA.

OR:

If the licensee cannot confirm that all internal events modeling updates performed to resolve F&Os that could impact fire risk were incorporated into the FPRA, then explain how the licensee will ensure that all internal events modeling updates performed to resolve F&Os that could impact fire risk are incorporated into the FPRA prior to implementation of the RICT program.

OR:

Explain how all the internal events modeling updates performed to resolve internal event F&Os have an inconsequential impact on the RICT calculations contribution from FPRA.

Q-24 Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty Confirm whether the licensee used any reduced transient heat release rates (HRRs) below the bounding 98 percent HRR of 317 kilowatts (kW) from NUREG/CR-6850 (ADAMS Accession Nos. ML052580075 and ML052580118). If yes, then EITHER:

Demonstrate that using reduced transient HRRs has an insignificant impact on this application.

OR:

Justify the use of the reduced HRRs, including:

• Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

• A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR, including how No.

Requested Documents (D) and Audit Questions (Q) location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location.

• The results of a review of records related to compliance with the transient combustible and hot work controls.

Q-25 Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty

a. Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents).

b. If the treatment of sensitive electronics for the FPRA includes deviations from FAQ 13-0004, then:

EITHER:

Identify the deviations, and justify (e.g., through a sensitivity calculation) that the treatment of sensitive electronics has no consequential impact on the RICT calculations.

OR:

Identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09, Revision 0-A, prior to implementation of the RICT program.

Q-26 10/25 Update:

Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty

a. Confirm and provide the minimum joint HEP value assumed in the FPRA.

b. EITHER:

If the FPRA used a minimum joint HEP (JHEP) value of less than 1E-5, then demonstrate (e.g., through a sensitivity study) that the minimum JHEP value(s) used have an inconsequential impact on the RICT application. If a sensitivity study is performed, then provide a description of and the quantitative results from the sensitivity study. Describe the process that will ensure that the impact of JHEP values below the thresholds used in future PRA model revisions remains minimal.

OR:

If the licensee cannot justify that the minimum JHEP value has an inconsequential impact on the application, then:

• Confirm that each FPRA JHEP value below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 (ADAMS Accession No. ML051160213) lower value guideline (i.e., using such criteria as the dependency factors identified in NUREG-1921 (ADAMS Accession No. ML12216A104) to assess level of dependence). Provide an estimate of the number of these joint HEP values below 1E-5, discuss the range of values, and provide at least two different examples where this justification is applied. Describe how JHEPs below the thresholds will be tracked as the PRA models evolve.

• If the licensee cannot justify JHEP values used in the FPRA below 1E-5, then identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09-A, prior to the implementation of the RICT program.

No.

Requested Documents (D) and Audit Questions (Q)

Q-27 Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty Describe how fire propagation outside of well-sealed motor control center cabinets greater than 440 Volts (V) is evaluated. If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, then provide justification for using this approach.

Q-28 Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty Provide a sensitivity study or other justification demonstrating that assigning weighting factors of 50 per the guidance in FAQ 12-0064 has an inconsequential impact on the RICT calculations.

Q-29 Refer to Background Document Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty Many plants have Unit 1 and 2 adjoined and, thus, have common areas. For these plants, the risk contribution from fires originating in one unit must be addressed for impacts to the other unit given the physical proximity of the other unit and common areas. Therefore, confirm whether Units 1 and 2 have common areas and shared systems, and if yes, then:

a. Explain how the risk contribution of fires originating in one unit is addressed for the other unit given the impacts from the physical proximity of equipment and cables in one unit to equipment and cables in the other unit. Include identification of locations where fire in one unit can affect components in the other unit.

b. Explain how the FPRA addresses contributions of fires in common areas that can impact components in both units, including the risk contribution of such scenarios. If any such scenarios are not addressed in the FPRA, then provide justification that their exclusion does not impact this application.

Q-30 Refer to Background Document LAR Enclosure 9, Table E9-2, states that the HRA for the fire PRA was performed using industry consensus modeling approaches but does not cite use of NRCs most current guidance on fire HRA, NUREG-1921. It is not clear if the licensee considered this guidance in its HRA. Therefore:

a. Confirm whether the licensee considered the guidance in NUREG-1921 to perform the HRA for the fire PRA. If it did, then describe any deviations from the guidance that could be characterized as potential key assumptions or sources of modeling uncertainty.

b. If in response to part (a) deviations from the guidance in NUREG-1921 could be characterized as assumptions or sources of modeling uncertainty, then justify that this modeling uncertainty has an inconsequential impact on the application (e.g., by performing a sensitivity study). If it cannot be determined that this modeling uncertainty has an inconsequential impact on the estimated RICTs, then identify programmatic changes to compensate for this uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty).

Q-31 Refer to Background Document Dispositions of Fire PRA Model Assumptions and Sources of Uncertainty EITHER:

Identify the systems or components that are assumed to be always failed in the fire PRA (or are not included in the fire PRA) caused by a lack of known cable routing and justify that this assumption has an inconsequential impact on the RICT calculations.

No.

Requested Documents (D) and Audit Questions (Q)

OR:

If it cannot be determined that the cited assumption has an inconsequential impact on the estimated RICTs, then identify programmatic changes to compensate for this uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty).

Q-32 Refer to Background Document Determination of Seismic CDF Penalty Justify that using a plant-level high confidence low probability of failure (HCLFP) capacity of 0.3g (acceleration due to gravity) is sufficient for estimating a bounding seismic CDF for this application given that there are plant SSCs with a HCLFP capacity of 0.21g. Include discussion of the cited outliers that were identified in the Susquehanna Individual Plant Examination of External Events (IPEEE) and the resolutions of those outliers for this application.

11/2/21 update: Include: (1) discussion on Group 1 outliers; (2) explanation of the commitments referenced in the Susquehanna IPEEE; (3) confirmation of actions taken for implementation of the commitments; and (4) resolution of any outliers for which the commitments were not performed.

Q-33 Refer to Background Document Determination of Seismic CDF Penalty Confirm whether there are outliers that include SSCs with a lower HCLPF capacity than 0.21g and whose failure can impact seismic CDF. If yes, then justify that using the lower seismic capacity for those components has an inconsequential impact on the RICT calculations.

Q-34 Refer to Background Document Determination of Seismic LERF Penalty Clarify how the percent seismic CDF contributions by sequence presented in the second, third, and fourth columns of LAR Table E4-3 and the seismic conditional large early release probability (CLERP) estimates for Susquehanna presented in the fifth column of the table were combined to determine the final sequence-weighted seismic CLERP values presented in the last row of the table (i.e., 0.36, 0.45, and 3.3E-2) for the three data sources.

Q-35 see Q-55 Refer to Background Document Determination of Seismic LERF Penalty Justify why using the average of the large difference presented in LAR Table E4-3 in the percent contribution of seismic CDF scenarios that go directly to core damage for Limerick Generating Station (LGS) and LaSalle County Station (LCS) is sufficient to support estimation of the LERF seismic penalty for Susquehanna, given that Susquehanna and Limerick use the GE Type 4 reactor while LCS uses the GE Type 5 reactor. Explain why the percent contribution of this scenario to the total seismic CDF for Susquehanna is not more like LGS (i.e., 33 or 43 percent) than it is like LCS

(<1 percent).

11/2/21 update: include: Revised seismic CLERP value based (1) on Farley method, or (2) Limerick data only or other plant data supported by justification using Susquehanna-specific design information for the applicability of the seismic data used.

No.

Requested Documents (D) and Audit Questions (Q)

Q-36 see Q-55 Refer to Background Document Determination of Seismic LERF Penalty Justify that using seismic information about the percent contribution of seismic CDF to the total from older seismic studies for LGS and LCS provide a sufficient basis for estimating the seismic LERF penalty for this application. Discuss the review of more current hazard or plant design information with the potential to impact the seismic accident scenario CDF contributions to the total seismic CDF. Discuss site and plant differences that could impact the seismic accident scenario CDF contributions to the total seismic CDF.

Q-37 Refer to Background Document External Flooding Identify and justify the mechanism that will be used to ensure that the watertight doors will be closed during a flood event to prevent impact on risk significant equipment.

11/2/21 update: include: (1) clarification of whether actions to close Diesel Generator (DG) Building doors and install panels are credited against local intense precipitation, (2) clarification that the entry condition to the severe weather/natural phenomena response procedure also includes non-tropical heavy rain and wind, and (3) clarification that a list of doors was not provided in the LAR.

Q-38 In the LAR, Enclosure 12, Risk Management Action (RMA) Examples, the licensee stated that multiple example RMAs may be considered during a RICT program entry to reduce the risk impact and ensure adequate defense-in-depth. Provide a list of RMAs that will to be considered during the implementation of RICT program relating to the following TS Conditions.

(a) TS 3.8.1, Condition C (b) TS 3.8.4, Condition B (c) TS 3.8.4, Condition C (d) TS 3.8.7, Condition A (e) TS 3.8.7, Condition B (f) TS 3.8.7, Condition C (g) TS 3.8.7, Condition D Q-39 For TSTF-505, the design success criterion is a minimum set of the remaining required equipment that has the capacity and capability to provide the TS safety function. In Table E1-1 of Enclosure 1 of the LAR, the licensee stated that the design success criteria for TS 3.8.1, Condition C (two offsite circuits inoperable), are two offsite circuits. Explain, how, with both required offsite circuits inoperable, can an offsite circuit provide the capacity and capability to safely shut down the reactor and maintain it in safe condition during and after a design basis accident.

Q-40 For TSTF-505, the design success criterion is a minimum set of the remaining required equipment that has the capacity and capability to provide the TS safety function. In Table E1-1 of Enclosure 1 of the LAR, the licensee describes the direct current (DC) power subsystems design for the design success criteria for TS 3.8.4, Conditions A, B, and C.

a. In the design success criteria for TS 3.8.4, Conditions A, B, and C, the licensee stated that the design success criterion is dependent upon the load supplied.

Explain the potential load supplied, as it is not defined in the design success criteria.

b. For TS 3.8.4, Conditions A, B, and C, the licensee states that one of two subsystems is the design success criterion and that four 125-volt DC (VDC) subsystems are available. How many 125-VDC subsystems are required to provide the safety function?

No.

Requested Documents (D) and Audit Questions (Q)

c. For TS 3.8.4, Conditions A, B, and C, the licensee states that one of two subsystems is the design success criterion and that two 250-VDC subsystems are available. How many 250-VDC subsystems are required to provide the safety function?

Q-41 In Table E1-1 of Enclosure 1 of the LAR, the licensee stated that the design success criterion for TS 3.8.7, Conditions A and B, is one subsystem and the design success criterion for Unit 2 TS 3.8.7, Conditions C and D is one of two divisions. Provide the definition of subsystems and divisions in the design success criteria column of Table E1-1 and the difference between these two terms. Also provide the definitions of channel, load group, and subgroup load.

Q-42 10/13/21 Update:

In Attachment 1 of the LAR, for Unit 2 TS 3.8.7, Conditions C and D, the licensee states that some components required by Unit 2 receive power through Unit 1 electrical power distribution subsystems. In Section 8.3.1.11.1 of the updated final safety analysis report (ADAMS Accession No. ML21294A245), the licensee states,

[T]here are no Unit 2 specific loads energized from the Unit 1 AC Distribution System.

a. Discuss which Unit 1 electrical power distribution subsystems provide power to Unit 2 components, and discuss which Unit 2 components receive power from Unit 1 electrical power distribution subsystems, especially, the configuration of the distribution system (e.g., buses, crossties) associated with the common load between the two units. In addition, the staff requests the licensee to confirm whether the crossties (if any) are explicitly modeled in the PRA. NRC also requests a simplified diagram showing distribution from Unit 1 to common loads and from common loads to Unit 2 equipment (from the DG to the power supplies and to common equipment).

Discuss whether there are any crossties between the two units to common loads.

Include list of common loads from a system level.

b. Additionally, in Table E1-1 of Enclosure 1 of the LAR, for Unit 2 TS 3.8.7, Conditions C and D, the licensee states, SSCs are modeled consistent with the TS scope and so can be directly included in the Real-Time Risk tool for the RICT Program. The staff requests the licensee to identify all shared equipment between Units 1 and 2 and Confirm that the shared equipment is explicitly components between Unit 1 and Unit 2 are modeled in the PRA.

c. Provide the results of a run the model for Unit 2 TS 3.8.7, Conditions C and D, regarding that shows the shared equipment components between Unit 1 and Unit 2 in the model, how are common loads set up in the model? Verify whether there any manual actions required and, if yes, verify whether they are credited in the PRA.

d. Assuming a station loss of offsite power (LOOP) concurrent with a LOCA in one unit, the staff requests the licensee to provide the distribution system configuration for safe shutdown of both units. If any crossties are being used, the staff requests the licensee to provide the discussion of the operator action.

Q-43 Table E1-2 of Enclosure 1 of the LAR provides the RICT estimates for TS Conditions in the scope of the RICT program.

a. Provide the high and low estimated RICT values for TS 3.8.1.A, TS 3.8.1.B, TS 3.8.1.D, TS 3.8.7.C (Unit 2), and TS 3.8.7.D (Unit 2).

b. Provide the possible ranges of RICT values of these TS conditions, considering the best and the worst configurations that could be anticipated.

Q-44 The licensee provided the engineered safeguard system bus load List for Unit 1 and Unit 2 in response to NRC audit request D-19. Discuss how many distribution No.

Requested Documents (D) and Audit Questions (Q) subsystems and buses are required for safety-related equipment to perform the safety function.

Q-45 (Follow-up to Q-41) Regarding TS 3.8.7, Condition C, One Unit 1 AC electrical power distribution subsystem inoperable, and TS 3.8.7, Condition D, Two Unit 1 AC electrical power distribution subsystems on one division inoperable for performance of Unit 1 Surveillance Requirement 3.8.1.19, the design success criteria for TS licensee to explain why the design success criteria are divisions and not subsystems. If the licensee changes the design success criteria based on the response to this question, then the staff requests a revision to Table E1-1 that reflects the change.

Q-46 The NRC staff requests the licensee to discuss how many safety buses are required for site LOOP concurrent with a LOCA in one unit.

Q-47 Regarding Table E1-4, Conditions 2.d and 2.e, and Table E1-5, Function 1.d, the staff requests the licensee to clarify the Tables regarding the accident and transient information.

Q-48 The staff requests the licensee to elaborate on diversity (defense-in-depth) of Function 3.d in TS Table 3.3.5.1-1, including discussion of manual actions.

Q-49 Regarding Table E1-9, for Function 3.e, Table E1-9 states automatic initiation. The staff requests the licensee to confirm that, if the only diversity identified in the Diverse Instrumentation column in tables E1-4 to E-12 and TS 3.3.8.1 is Manual, then such manual actions are credited in the PRA model and prescribed in operation manuals and procedures.

Q-50 The staff requests the licensee to elaborate on diversities (defense in depth) of all functions in TS Table 3.3.8.1-1.

Q-51 Attachment 2 includes markups of the TS to support the proposed implementation of a RICT program at Susquehanna Unit 1 and Unit 2, respectively. Address the following inquiries and observations relative to these markups:

a. The Unit 1 TS Table 3.3.5.1-1 pages are marked to reduce the table from 6 pages to 5 pages. Provide justification for this change.

b. The proposed administrative controls for the RICT Program in TS 5.5.16, paragraph e of Attachment 2 of the LAR were based on the TS markups of TSTF-505, Revision 2 for Susquehanna Unit 1 and Unit 2. The NRC staff recognizes that the model safety evaluation for TSTF-505, Revision 2 contains improved phrasing for the administrative controls for the RICT Program in TS 5.5.16, paragraph e, namely the phrasing approved for use with this program instead of used to support this license amendment. In lieu of the original phrasing in paragraph e of TS 5.5.16, discuss whether the phrases used to support Amendment ### or, as discussed in the TSTF-505 model safety evaluation, approved for use with this program would provide more clarity for this paragraph. The proposed administrative controls in TS 5.5.16, paragraph e include the phase, this license amendment. In lieu of the phrase, this license amendment, discuss whether the phrases Amendment ### or, as discussed in the TSTF-505 model safety evaluation, this program would provide more clarity for the paragraph.

c. TS Required Actions 3.7.1.B.1, 3.7.2.B.1, and 3.7.2.C.1 have existing temporarily extended completion times for specific events. The licensee has proposed to also include these Required Actions in the RICT program. Discuss how the temporary extensions and the RICT would be implemented.

Q-52 The staff requests the licensee to upload an updated version of Table E-1 under this question.

No.

Requested Documents (D) and Audit Questions (Q)

Q-53 [Placeholder for a request that the NRC staff determined was no longer needed.]

Q-54 Section C.1.4 of RG 1.200 states that the base PRA (i.e., the model of record) is to represent the as-built, as-operated plant to the extent needed to support the application. The licensee is to have a process that identifies updated plant information that necessitates changes to the base PRA model. In response to an event involving an open-phase condition (OPC) at the Byron Generating Station on January 30, 2012, the NRC issued Bulletin 2012 01.[1] As part of the initial voluntary industry initiative for mitigation of the potential for the occurrence of an OPC in electrical switchyards,[2]

licensees have made the addition of an open-phase isolation system (OPIS). In Staff Requirements Memorandum (SRM)-SECY 16-0068,[3] the NRC staff was directed to ensure that licensees have appropriately implemented OPIS and that licensing bases have been updated accordingly. NRC staff closed out BL 2012-01 for Susquehanna via letter dated December 6, 2021.[4] From the revised voluntary initiative[5] and resulting industry guidance on estimating OPC and OPIS risk in NEI 19-02,[6] it is understood that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design. Therefore, OPC could impact the RICTs for some TS LCO conditions within the scope of the RICT program (e.g., conditions associated with TS LCOs 3.8.1, 3.8.7, and 3.8.4). Considering these observations, provide the following information:

a) For Susquehanna, discuss the evaluation of the risk impact associated with OPC events including the likelihood of OPC-initiated plant trips and the impact of those trips on PRA-modeled SSCs. Also, discuss the functionality of the open phase detection system installed at Susquehanna and operator actions needed to operate or respond to the system.

b) Clarify whether any installed equipment and associated operator actions are credited in the PRAs that support this application. If equipment and associated operator actions are credited, then provide the following information:

i. Describe the equipment and associated actions that are credited in the PRA models.

ii. Describe the impact that this treatment, if any, has on key assumptions and sources of uncertainty for the RICT program.

iii. Discuss HRA methods and assumptions used for crediting alarm manual response.

iv. Discuss how OPC-related scenarios are modeled for non-internal event scenarios such as internal floods, fire, and seismic.

v. Regarding inadvertent actuation of the open phase detection system open phase isolation:

• Explain whether scenarios regarding inadvertent actuation of the system, if applicable, are included in the PRA models that support the RICT calculations.

• If inadvertent actuation scenarios are not included in the PRA models, then provide justification that the exclusion of this inadvertent actuation does not impact the RICT calculations.

c) If OPC and the open phase detection system and ability to manually mitigate an open phase condition are not included in the application PRA models, then provide justification that the exclusion of this failure mode and mitigating system does not impact the RICT calculations.

d) As an alternative to Part (c), propose a mechanism to ensure that OPC-related scenarios are incorporated into the application PRA models prior to implementing the RICT program.

1 U.S. NRC Bulletin 2012-01, Design Vulnerability in Electric Power System (ADAMS Accession No. ML12074A115).

No.

Requested Documents (D) and Audit Questions (Q) 2 Anthony R. Pietrangelo to Mark A. Satorius, ltr re: Industry Initiative on Open Phase Condition -- Functioning of Important-to-Safety Structures, Systems and Components (SSCs), dated October 9, 2013 (ADAMS Accession No. ML13333A147).

3 U.S. NRC SRM-SECY-16-0068, Interim Enforcement Policy for Open Phase Conditions in Electric Power Systems for Operating Reactors, dated March 9, 2017 (ADAMS Accession No. ML17068A297).

4 ADAMS Accession No. ML21335A422 5 Doug True to Ho Nieh, ltr re: Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ADAMS Accession No. ML19163A176).

6 Nuclear Energy Institute (NEI) 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, Revision 0, April 2019 (ADAMS Accession No. ML19122A321).

Q-55 Provide a copy of the updated analysis for Seismic CDF and LERF estimate for the RICT Program based on discussion at the audit meeting held on December 1, 2021.

NRC Background Document No. 1 for Audit Questions Q2 through Q37 August 9, 2021 AUDIT QUESTIONS SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 License Amendment Request To Adopt Risk-Informed Extended Completion Times Probabilistic Risk Assessment Licensing Branch A (APLA)

PRA Acceptability and Risk-informed Approach Open Internal Events PRA Facts and Observations AUDIT QUESTION Q-2 Regulatory Guide (RG) 1.200, Revision 2 (ADAMS Accession No. ML090410014) provides guidance for addressing probabilistic risk assessment (PRA) acceptability. RG 1.200, Revision 2, describes a peer review process using 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, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of 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 05-04/07-12/12-06 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).

LAR Enclosure 2, Table E2-1 presents the disposition for F&O 1-18, which remains open after the internal events full-scope peer review performed in October 2012 and Independent Assessment to close F&Os in April 2018 and September 2020. The disposition indicates that Susquehanna plans to review and update the list of internal events PRA modeling assumptions and sources of uncertainty based on the disposition of the 2020 F&O closure review team and provide dispositions that are specific to this application prior to implementation of TSTF-505.

NRC staff notes that review and update of the list of internal events PRA modeling assumptions and sources of uncertainty could lead to identification of issues that are found to be key and, therefore, could have an impact on the application.

EITHER:

Confirm that F&O 1-18 has been resolved by showing that Susquehanna has (1) reviewed and updated the list of internal events PRA modeling assumptions and sources of uncertainty based on the disposition of the 2020 F&O closure review team, (2) provided dispositions that are specific to this application, and (3) addressed, in accordance with NEI 06-09, Revision 0-A, any assumptions or sources of uncertainty determined to be potentially key to this application (e.g., performed a sensitivity study demonstrating that they have an inconsequential impact on the risk-informed completion time (RICT) calculations, or identified programmatic changes to compensate for this modeling uncertainty). Also, confirm whether Report EC-RISK-0056, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications (dated February 8, 2021), which predates the TSTF-505 LAR dated April 8, 2021, represents the review update of the internal events PRA uncertainty analysis.

OR:

If the licensee cannot confirm that F&O 1-18 has been resolved per the above, then explain how the licensee will ensure (e.g., via a license condition or an implementation requirement), prior to implementation of the Risk-Managed Technical Specifications (RMTS) program, that it will: (1) review and update the list of internal events PRA modeling assumptions and sources of uncertainty, (2) update the associated dispositions that are specific to this application, and (3) fully address any impacts on the RMTS program (including those determined to be key for this application) in accordance with NEI 06-09, Revision 0-A.

Dispositions of PRA Model Assumptions and Sources of Uncertainty The NRCs safety evaluation for NEI 06-09, Revision 0 (ADAMS Accession No. ML122860402),

specifies that the LAR should identify key assumptions and sources of uncertainty and to assess and disposition each as to their impact on the application of RMTS. Section 7 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.

of the LAR discusses the process for identifying key assumptions and sources of uncertainty from plant-specific and generic industry sources that appear to have the potential to impact the application and discusses the bases for screening modeling uncertainties from further consideration. The LAR states that this screening process includes identifying the approach used (e.g., consensus approach or other applicable guidance) and the level of detail included in the PRA model. LAR Enclosure 9 provides dispositions for candidate key assumptions and sources of uncertainty not screened in Tables E9-1, E9-2, and E9-3 for this application. For the internal events PRA, these dispositions discuss the results of sensitivity analyses to show the impact of the modeling uncertainty on RICT calculations. However, certain aspects of these sensitivity analyses that are important to the conclusions of the submittal are not clear to NRC staff.

AUDIT QUESTION Q-3 LAR Enclosure 9, Table E9-1 for the internal events PRA, states for most sources of uncertainty that sensitivity studies (the results of which are not presented in the LAR) show the RICT values are not significantly impacted by the modeling uncertainty and, therefore, no new risk management actions (RMAs) would be identified. The LAR states for these sources of uncertainty that the sensitivity analyses were performed for a select group of technical specifications. The LAR indicates that for one source of uncertainty (i.e., vapor suppression capability) the modeling impact was determined for six selected RICT estimates. The NRC staffs review of the uncertainty analysis report for risk-informed applications (EC-RISK-0056) found that for some sources of uncertainty, just three or four RICT estimates were performed for these sensitivity studies. The bases for selecting the plant configurations (i.e., TS LCO conditions) included in the sensitivity studies is not clear. Also, it is not clear how general conclusions can be reached about the impact of a modeling uncertainty based on what appears to be a small set of RICT estimates for each sensitivity study.

Therefore:

a.

Describe how the plant configurations (i.e., TS LCO conditions) were chosen in the sensitivity studies to assess the impact of potential key assumptions and sources of uncertainty. Also include in this discussion the bases for selecting these plant configurations.

b.

Justify how the process described in part (a), above, is sufficient to conclude that, based on the sensitivity study RICT estimates, the impact of the associated modeling uncertainty on the RICT calculations is inconsequential.

AUDIT QUESTION Q-4 LAR Enclosure 9, Table E9-1 states that concerning the room heatup calculations, operator failure to open the Engineered Safeguard Service Water (ESSW) pumphouse door is a source of modeling uncertainty. The LAR explains that when there are no operating cooling fans and all four ESSW pumps are running, the pumps should still be able to complete their 24-hour mission if the pump house doors are opened within two hours of pump start. The NRC staffs review of the uncertainty analysis report for risk-informed applications (EC-RISK-0056, Section 3.1) found that for the sensitivity case addressing this modeling uncertainty, the fire PRA core damage frequency (CDF) increased by 71 percent if the operator failure probability to open the pump house doors (i.e., human failure event 016-N-N-VENT-O) was increased by a factor of ten. This sensitivity study also shows that the RICT estimates for two LCO conditions (i.e., 3.7.1.B and 3.8.1.B) are significantly decreased in the sensitivity case. LAR Enclosure 9, Table E9-1 indicates this issue has two aspects: (1) the uncertainty associated with the fragility of components to loss of cooling, and (2) the uncertainty associated with the assumptions about the rate of room heatup and the timing of the component failure caused by the heatup.

However, the LAR then concludes:

This is not a realistic assumption given the feasibility of the operator action in question and use of this assumption would result in overly-conservative RICT estimates.

Therefore, this does not represent a key source of uncertainty for the RICT application.

It is not clear to NRC staff how the unfavorable sensitivity study results indicating that the modeling uncertainty impacts the RICT calculations justify selecting an inappropriate sensitivity case. Therefore:

EITHER:

Report the results of a sensitivity study in which the increase in the operator failure probability in the sensitivity case is set low enough that it is not unrealistic but high enough that it tests the modeling uncertainty to demonstrate that this modeling uncertainty is not key for the RMTS program and has an inconsequential impact on the RICT calculations. Also, describe this sensitivity study and justify the appropriateness of the selected operator failure probability used in the sensitivity case. Provide the bases for the chosen plant configurations (i.e., TS LCO conditions) in this sensitivity study.

OR:

Describe the programmatic changes to compensate for this modeling uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). This discussion should also identify the TS LCO conditions in scope of RMTS for which the RICT calculations are impacted by this uncertainty and discuss how the RICTs are impacted (e.g., describe and provide the results of applicable sensitivity studies). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) provide RMA examples that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

OR:

Provide a detailed justification (e.g., propose an implementation item to update the PRA to address this modeling uncertainty and discuss how this update addresses this uncertainty; describe and provide the results of a different type of sensitivity study) that this modeling uncertainty does not need to be addressed in the RMTS program as required by Section 2.3.4 of NEI 06-09-A.

Credit for FLEX Equipment and Actions The NRC memorandum, 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, dated May 30, 2017 (ADAMS Accession No. ML17031A269),

provides the NRCs staff assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014).

Regarding equipment failure probability, the NRC staff concludes (Conclusion 8) the following in its memorandum dated May 30, 2017:

The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.

Regarding human reliability analysis (HRA), NEI 16-06 Section 7.5 recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections 7.5.4 and 7.5.5 of NEI 16-06 describe actions to which the current HRA methods cannot be directly applied (e.g., debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses) and those complex actions that require many steps over an extended period (e.g., multiple personnel and locations, evolving command and control, and extended time delays). The NRC staff concludes (Conclusion 11) the following in its memorandum dated May 30, 2017:

Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, HEPs [human error probabilities] associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.

Regarding uncertainty, Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties that could potentially impact the results of a RICT calculation. NEI 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate RMAs, including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in PRA modeling of FLEX strategies related to the equipment failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application.

LAR Enclosure 2, Section 6 states that certain FLEX equipment is credited in the PRA. The LAR states that the incorporation of FLEX equipment into the PRA models was done using the same approach used for previous model revisions performed to reflect the as-built, as-operated plant. However, the LAR also states, [D]ue to limited reliability data for FLEX equipment, conservative adjustments were made to the generic reliability values for similar equipment.

The LAR states, Human Error Probabilities (HEPs) for FLEX components were evaluated with the same methodology used for all other HEPs in the Susquehanna PRA models. The LAR explains that sensitivity analyses were performed to demonstrate the FLEX modeling does not have a significant impact on the calculated RICT[s] and, therefore, RICT calculations are not sensitive to equipment reliability and HEP values. The LAR does not describe or present the results of those sensitivity studies, but an uncertainty analysis sensitivity case (EC-RISK-0056, Section 3.6) indicates that the RICT for LCO 3.8.1.C decreased by 37 percent when the FLEX portable equipment failure rates were increased by a factor of two. In general, NRC staff notes the significant challenges of modeling FLEX equipment and actions without sufficient industry data and without a consensus HRA approach to address unique aspects of FLEX actions.

Given the observations above, and the fact that a description of the sensitivity studies performed on the impact of crediting FLEX strategies and a presentation of those results are not provided in the LAR, it is not clear that the results of the sensitivity studies are sufficient to conclude the credited FLEX modeling has an inconsequential impact on calculated RICTs for the plant configurations allowed under the proposed TSTF-505 program. Therefore:

AUDIT QUESTION Q-5 Describe the FLEX strategies credited for the internal events (including internal flooding) and fire PRA models.

AUDIT QUESTION Q-6 Identify the FLEX equipment credited and whether that equipment is portable or permanently installed, and:

a. Discuss whether the credited FLEX equipment (regardless of whether it is portable or permanently installed) is similar to other plant equipment credited in the PRA (e.g.,

systems, structures, and components (SSCs) with sufficient plant-specific or generic industry data).

b. For credited FLEX equipment that is not similar to other plant equipment credited in the PRA:

i.

Discuss the data and failure rates used to support its modeling and provide the rationale for using the chosen data and any conservative adjustments that were made to the generic reliability values for similar equipment. Discuss whether the uncertainties associated with the parameter values are in accordance with the ASME/ANS PRA Standard, as endorsed by RG 1.200, Revision 2.

ii. Describe the sensitivity study performed to assess the impact of uncertainty associated with equipment failure probabilities on calculated RICTs and present the results of that study. Justify how the increase in equipment failure probabilities used in the sensitivity case constitutes bounding realistic estimates.

Also, discuss the bases for the chosen TS LCO conditions in the sensitivity study. Because the 30-day RICT back-stop condition could mask the impact of this uncertainty in the sensitivity study, discuss whether the RICTs for plant configurations involving more than one LCO entry (e.g., where the calculated RICTs are less than the 30-day backstop) are significantly impacted by this uncertainty.

iii. Discuss whether the uncertainty associated with equipment failure probabilities is a key source of uncertainty for the RMTS program. If this uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed in the RMTS program (e.g., programmatic changes such as identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) for those TS LCOs in LAR Enclosure 12 (Risk Management Action Examples) that are significantly impacted by this uncertainty, provide updated RMAs that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

AUDIT QUESTION Q-7

a. Identify the FLEX operator actions credited in the PRA and discuss whether any of these operator actions contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06.

b. For credited operator actions related to FLEX equipment that contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06:

i.

Describe the sensitivity study performed to assess the impact of uncertainty associated with FLEX HEPs (both the FLEX independent and dependent HEPs) on calculated RICTs and present the results of that study. Justify how the increase in the FLEX HEPs used in the sensitivity case constitutes bounding realistic estimates. Also, discuss the bases for the chosen TS LCO conditions in the sensitivity study. Because the 30-day RICT back-stop condition could mask the impact of this uncertainty in the sensitivity study, discuss whether the RICTs for plant configurations involving more than one LCO entry (e.g., where the calculated RICTs are less than the 30-day backstop) are significantly impacted by this uncertainty.

ii. Discuss whether the uncertainty associated with FLEX HEPs is a key source of uncertainty for the RMTS program. If this uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed in the RMTS program (e.g., programmatic changes such as identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A; and (2) for those TS LCOs in LAR 2, Risk Management Action Examples, that are significantly impacted by this uncertainty, provide updated RMAs that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty, and explain how these RMAs are expected to reduce the risk associated with this uncertainty.

AUDIT QUESTION Q-8 Section 1-2 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard, as endorsed by RG 1.200, Revision 2, 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 1-5.4 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard states that PRA upgrades shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard.

Given the challenges of modeling FLEX mitigation strategies, explain whether the review of the FLEX modeling was included in the last peer review of the PRA models. If it was not, then justify how the model changes associated with incorporating FLEX mitigating strategies does not constitute a PRA upgrade as defined in Section 1-2 of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2.

Impact of SOKC Uncertainty the RICT Calculations AUDIT QUESTION Q-9 RG 1.174 clarifies that 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.

However, point estimate PRA results are commonly calculated and reported, but these do not account for the state-of-knowledge correlation (SOKC) between nominally independent basic event probabilities. Mean values reflect the SOKC and are typically larger than point estimates, but require longer and more complex calculations. NUREG-1855, Revision 1 provides guidance on evaluating how the SOKC uncertainty impacts the comparison of the PRA results with the guideline values.

The SOKC uncertainty results for the internal events PRA in Notebook EC-RISK-0040 (Table 3.10-1) show the difference between the point estimate and mean (50,000 samples) core damage frequencies (CDFs) and large early release frequencies (LERFs) for Units 1 and 2 to be from 20% to 45%. The SOKC uncertainty results for the fire PRA in Notebook EC-RISK-1187 (Table 4-12) shows the difference between the point estimate and mean LERFs for Units 1 and 2 to be 15% and 18%, respectively. Notebook EC-RISK-0040 states, [t]he mean values using random samples, as expected, are only slightly higher than the quantification value, indicating results confidence. However, the stated differences between the point estimate and mean risk values appear to be large and could potentially impact the RICT calculations (i.e., RICTs calculated using mean CDF/LERF values could be significantly less than the RICTs calculated using point estimate CDF/LERF values).

Therefore:

a. Provide a summary of how the SOKC investigation was performed for the base Susquehanna PRA models used to support the RMTS application. Provide and discuss the results of this SOKC investigation and whether the SOKC uncertainty has a significant impact on the RICT calculations.

b. Provide and discuss the results of a comparison study between the RICT values calculated using point estimate risk versus mean risk for various LCO conditions in scope of RMTS. The LCO conditions selected for this comparison study should have a point estimate RICT less than 30 days (i.e., the 30-day backstop does not mask the comparison results) and are considered most likely to be impacted by the SOKC uncertainty (i.e., point estimate RICT versus mean RICT).

Provide the bases for the chosen LCO conditions in this comparison study. Also, provide the intermediate risk results from these RICT calculations (e.g., the CDFs and LERFs for the baseline case using point estimates and sensitivity case using mean values from the internal events, including internal flooding, and fire PRAs).

c. Based on the results above, provide a summary of how the SOKC will be addressed for RICT calculations during RMTS implementation (i.e., based upon the risk metrics to be considered), and explain how this process/approach is consistent with NUREG-1855, Revision 1.

Total Risk and Accounting for the SOKC AUDIT QUESTION Q-10 RG 1.174 provides the risk acceptance guidelines for total CDF (i.e., CDF < 1E-04 per year) and LERF (i.e., LERF < 1E-05 per year). Per RG 1.174 and Section 6.4 of NUREG-1855, Revision 1, for a Capability Category 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 probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly represented in the PRA models. In general, the point estimate CDF and LERF 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. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the SOKC is unimportant (i.e.,

the risk results are well below the risk acceptance guidelines).

LAR Enclosure 5, Section 2 states that the total CDF and LERF values presented in Table E5-1 for Susquehanna are point estimate values, which are likely lower than the mean CDF and LERF values. In addition to this, NRC staff notes that the current PRA models could potentially be updated in response to information requests (such as the response to requests on potentially unresolved fire PRA F&Os, request to update the fire PRA with resolutions used to close out internal events PRA F&Os, requests on the acceptability of fire PRA methods, and requests concerning FLEX modeling). NRC staff notes that for Susquehanna Units 1 and 2, the total LERF presented in the LAR is 6.9E-06 per year. Therefore, the total LERF could approach the RG 1.174 risk acceptance guideline of 1E-05 per year when the total mean LERF is used accounting for the SOKC and potential risk increases associated with model updates.

EITHER:

Demonstrate that the total risk for Susquehanna Units 1 and 2 is in conformance with RG 1.174 risk acceptance guidelines (i.e., CDF < 1E-04 and LERF < 1E-05 per year) after the total mean internal events (including internal flooding) and fire CDF and LERF values are calculated to account for the SOKC and for potential changes in risk due to any updates to PRA models performed in response to NRC staff requests. Identify the fire PRA parameters that are assumed to be correlated in the parametric uncertainty analysis of fire events (e.g., fire ignition frequencies, non-suppression probabilities, severity factors, spurious probabilities, fire human error probabilities), as well as the sources used for the associated uncertainty distributions (e.g., NUREG-2169, NUREG/CR-1278, NUREG/CR-7150, and EPRI HRA Calculator uncertainty distributions).

OR:

Explain how the licensee will ensure (e.g, via a license condition or implementation requirement) that, prior to implementation of the RMTS program: (1) the total mean internal events (including internal flooding) and fire CDF and LERF will be calculated to account for the SOKC and updates to PRA models performed in response to NRC staff requests; and (2) the updated total risk (including seismic risk) values are still in conformance with the RG 1.174 risk acceptance guidelines (i.e., CDF < 1E-04 and LERF < 1E-05 per year).

Consideration of Shared Systems in RICT Calculations AUDIT QUESTION Q-11 The Tier 3 assessment in RG 1.177 stipulates that a licensee should develop a program that ensures the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity.

LAR Attachment 1, Section 2.3 states that the Emergency Service Water (ESW) system is shared between the two units and consists of two loops, each of which is designed to supply 100 percent ESW requirements to both units and the common diesel generators (DGs) simultaneously. Besides the ESW and DGs, it is not clear whether other systems are shared between units. NRC staff notes that for certain events, such as dual unit events (e.g., loss of offsite power), it may be appropriate to only credit the shared systems for one unit. Therefore:

Explain whether shared systems are credited in the internal events (including internal flooding) and fire PRA models for both units that support the RICT calculations and, if so, then (1) identify those systems, and (2) either explain how the shared systems are modeled for each unit in a dual unit event demonstrating that shared systems are not over-credited in the PRA models, or if the PRA models do not address the impact of events that can create a concurrent demand for the system shared by both units, then justify that this exclusion has an inconsequential impact the RICT calculations.

Impact of Seasonal Variations AUDIT QUESTION Q-12 The Tier 3 assessment in RG 1.177 stipulates that a licensee should develop a program that ensures that the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09 and its associated NRC safety evaluation state that, for the impact of seasonal changes, either conservative assumptions should be made or the PRA should be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration. LAR Enclosure 8, Section 2 states that when there are seasonal dependences, the Real-Time Risk model addresses the average configuration of the plant and that plant-specific configurations will be evaluated as needed.

Explain how the impact of seasonal variations on the PRA modeling will be evaluated (as needed) during a RICT evolution and justify that this approach is consistent with the guidance in NEI 06-09 and its associated NRC safety evaluation.

In-Scope LCOs and Corresponding PRA Modeling The NRC safety evaluation for NEI 06-09 specifies that a LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. LAR, Table E1-1 identifies each TS LCO proposed for the RICT program, describes whether the systems and components involved in the TS LCO are implicitly or explicitly modeled in the PRA, and compares the design basis and PRA success criteria. For certain TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be conservatively represented using a surrogate event. For some LCO conditions, the LAR did not provide enough description for NRC staff to conclude that the PRA modeling will be sufficient for each proposed LCO Condition.:

AUDIT QUESTION Q-13 (STSB)

LAR Table E1-1 states for TS LCO 3.5.1 (ECCS - Operating), Condition D (HPCI System Inoperable) that both the design basis and PRA success criterion is one of one train (i.e., one HPCI pump). It appears that LCO 3.5.1, Condition D defeats the design basis success criterion and, therefore, represents a TS loss of function. TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493) does not authorize determination of a RICT when the condition represents a loss of TS function. Therefore, explain why LCO 3.5.1, Condition D does or does not represent a TS loss of function. Include clarification of the design basis success criteria for the High Pressure Coolant Injection (HPCI) system.

AUDIT QUESTION Q-14 LAR Table E1-1 states for TS LCO 3.5.1 (ECCS - Operating), Condition B (One LPCI pump in one or both LPCI subsystems inoperable), that the PRA success criteria are generally consistent with the design basis. However, the table also shows that for a LOCA in the bottom head, the PRA success criterion is one RHR pump in each division. This appears to be more stringent than the design basis success criterion (i.e., one of four LCPI pumps) and associated with a specific accident sequence.

Therefore, explain this apparent inconsistency and confirm whether the more stringent PRA success criterion is for one possible low likelihood event.

AUDIT QUESTION Q-15 LAR Table E1-1 states for TS LCO 3.7.2 (Emergency Service Water (ESW) System),

Conditions A, B, and C, that the success criteria are consistent with the design basis.

However, the table indicates that the design basis success criterion is one ESW pump in each loop, and the PRA success criterion is one of two subsystems. The design basis and PRA success criteria do not appear to be equivalent. Therefore, explain how the design basis and PRA success criteria are consistent based on the wording in Table E1-1. If the licensee cannot confirm that the design basis and PRA success criteria are consistent, then explain the basis for the difference and the justification for the PRA success criteria.

PRA Modeling of Instrumentation and Controls AUDIT QUESTION Q-16 Concerning the quality of the PRA model, NEI 06-09 states, RG 1.174, Revision 1, and RG 1.200, Revision 1 define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.

For several TS LCO Conditions listed in LAR Table E1-1, the table indicates that instrumentation and control (I&C) modeling in PRAs is insufficient to model the plant configuration and, therefore, the inoperability of the associated SSC (e.g., channel) will be modeled using a surrogate event. Accordingly, there appears to be general variability in the level of detail used to model the I&C systems creating some uncertainty about whether there is sufficient detail to support implementation of the proposed LCO conditions other than those specifically identified in LAR Table E1-1 that will be modeled using a surrogate event.

Therefore:

Explain how I&C equipment that is applicable or impacts the RICT calculations is modeled or 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., whether all channels of an actuation circuit are modeled), (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.

Uncertainty Associated with Digital Instrumentation and Control Modeling AUDIT QUESTION Q-17 Section 2.3.4 of NEI 06-09 states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties that could potentially impact the results of a RICT calculation.

Regarding digital I&C, there is a lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed applications. In addition, known modeling challenges exist, such as the lack of industry data for digital I&C components, the difference 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. However, it is not clear to NRC staff whether the licensee credited digital systems in the PRA models that will be used in the RICT program or whether this modeling can impact the RICT calculations. Therefore:

EITHER:

Describe and provide the results of a sensitivity study performed for each digital system modeled in the PRA demonstrating that the uncertainty associated with PRA modeling the digital I&C systems has inconsequential impact on the RICT calculations.

OR:

Identify the LCOs impacted by digital I&C system modeling and 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 requires additional RMAs.

PRA Update Process AUDIT QUESTION Q-18 Section 2.3.4 of NEI 06-09 specifies, criteria 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 a 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 a plant change meets specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the plant 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.

Performance Monitoring AUDIT QUESTION Q-19 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. NEI 06-09 was approved by the NRC on May 17, 2007 (ADAMS Accession No. ML071200238). The NRC safety evaluation (SE) for NEI 06-09, states, [t]he impact of the proposed change should be monitored using performance measurement strategies. NEI 06-09 considers the use of NUMARC 93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (ADAMS Accession No. ML18120A069), as endorsed by RG 1.160, Revision 4 (ADAMS Accession No. ML18220B281), for the implementation of the Maintenance Rule.

NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.

Furthermore, Section 2.3 of LAR Attachment 1 states, In addition, the NEI 06-09-A methodology satisfies the five key safety principles specified in RG 1.177, Revision 0, relative to the risk impact due to the application of a RICT.

NRC staff position C.3.2 provided in RG 1.177 for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period of time. It is unclear how the licensees RMTS process captures performance monitoring for the SSCs within-scope of the RMTS program. Therefore:

EITHER:

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

OR:

Describe the approach or method used by Susquehanna for SSC performance monitoring, as described in Regulatory Position C.3.2 of RG 1.177, for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative or quantitative), along with the appropriate risk metrics, and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC SE for NEI 06-09.

Probabilistic Risk Assessment Licensing Branch B (APLB)

Fire PRA Questions Open Fire PRA F&Os Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014) provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA Standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-Sa-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of 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 Nuclear Energy Institute (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 (F&Os) (ADAMS Package Accession No. ML17086A431), which was accepted by the NRC in a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427).

LAR Enclosure 2, Table E2-2 presents the dispositions for F&Os that remain open after the fire events full-scope peer review performed in February 2018 and independent assessment to close F&Os in June 2019. A few of the dispositions do not appear to completely resolve the F&O and, therefore, more information is needed to conclude that the open F&Os do not have a consequential impact on the RICT program.

AUDIT QUESTION Q-20 The LAR indicates that the peer review teams resolution for fire PRA F&O 1-9 was to incorporate Multiple Spurious Operations (MSO) scenarios 2al and 5f. The Susquehanna disposition (on the impact to the TSTF-505 program) for F&O 1-9 indicates that the modeling for MSO scenario 5f has been incorporated into the fire PRA, but the modeling for MSO scenario 2aj has not been incorporated. MSO scenario 2aj is not cited in F&O 1-9, and its risk significance is not discussed in the LAR, but the LAR indicates that it needs to be incorporated into the fire PRA as part of the PRA maintenance process. The disposition explains that MSO scenario 2al is not required to be modeled given the configuration of the Condensate Storage Tank and lack of a Condensate Return Tank at Susquehanna. It is not clear from the LAR how risk significant MSO scenario 2aj is or whether it could impact the RICT calculations for certain plant configurations. Therefore:

EITHER:

Justify that the incomplete fire PRA modeling associated with MSO scenario 2aj has an inconsequential impact on the calculated RICTs.

OR:

If the licensee cannot justify that the incorporation of MSO scenario 2aj has an inconsequential impact on the calculated RICTs, then describe how the licensee will ensure that MSO scenario 2aj is incorporated into the fire PRA model prior to implementation of the RICT program.

AUDIT QUESTION Q-21 The Susquehanna disposition of the impact to the TSTF-505 program from F&O 5-4 states that there is one credited cable that remains unlocated, but that [it] is very unlikely that one cable will have an impact of the overall FPRA results. The disposition from the F&O closure team indicates that if the cable location cannot be ascertained, then associated PRA components should be assumed failed and a sensitivity study should be performed in this case. The Susquehanna disposition for F&O 5-4 does not indicate that a sensitivity study assuming failure of the PRA components supported by the unlocated cable was performed. It is not clear to NRC staff which PRA components are supported by the cited cable and how risk significant the cable is, and whether the treatment could impact the RICT calculations for certain plant configurations. Therefore:

EITHER:

Discuss the PRA components that are supported by the unlocated cable, and justify (preferably using the sensitivity recommended by the peer reviewers) that the treatment of the unlocated cable has an inconsequential impact on the calculated RICTs.

OR:

If the licensee cannot justify that the treatment of the unlocated cable has an inconsequential impact on the calculated RICTs, then explain how the licensee will either ensure that the cable is located and properly modeled in the fire PRA prior to implementation of the RICT program or identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09-A.

AUDIT QUESTION Q-22 F&O 7-1 states that inadequate discussion and justification has been provided in the PRA documentation for crediting active partitions in the fire PRA. The Susquehanna disposition on the impact to the TSTF-505 program for F&O 7-1 states that PRA documentation has been updated to show that the fire barrier rating is indicated in the barrier name. However, the disposition also states that justification for crediting active partitioning elements has not been added. The disposition states that this documentation deficiency does not impact the TSTF-505 application. However, it is not clear how active partitioning elements can be credited if the justification has not been provided and reviewed. Therefore:

EITHER:

Identify the active partitions that have been credited in the FPRA and provide justification that they will perform reliably in the accident scenarios for which they are credited.

Include discussion of systems that rely on supporting systems, such as alternating current power or a water supply, to perform their functions.

OR:

Explain how removal of the credit taken for active partitions has an inconsequential impact on the RICT calculations.

Update of Fire PRA with Internal Event F&O Resolutions AUDIT QUESTION Q-23 RG 1.200, Revision 2 (ADAMS Accession No. ML090410014) provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS RASa-2009, "Addenda to ASME/ANS RASa-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os.

LAR Enclosure Section 3 states that multiple internal event PRA model adjustments were performed to resolve the F&Os identified from the 2012 full-scope internal events PRA peer review ahead of Independent Assessments performed in April 2018 and September 2020 to close F&Os using Appendix X to NEI 05-04, 07-12, and 12-13. LAR Enclosure Section 4 states that the last full-scope peer review of the FPRA was performed in February 2018 which is before the internal events PRA F&O closure review in April 2018. The LAR then states that the fire PRA F&O closure review was performed in June 2019 which is before the final internal events PRA F&O closure review in September 2020. Accordingly, it appears that modeling updates to the internal events PRA to resolve F&Os could have occurred after the model of record FPRA was finalized. Given that internal events PRA provides the modeling foundation for the FPRA and the observations above, it is not clear to NRC staff whether F&O resolutions made to the internal events PRA to close F&Os that could impact the FPRA were incorporated into the FPRA. Therefore:

EITHER:

Confirm that all internal events modeling updates performed to resolve internal event F&Os that could impact fire risk were incorporated into the FPRA.

OR:

If the licensee cannot confirm that all internal events modeling updates performed to resolve F&Os that could impact fire risk were incorporated into the FPRA, then explain how the licensee will ensure that all internal events modeling updates performed to resolve F&Os that could impact fire risk are incorporated into the FPRA prior to implementation of the RICT program.

OR:

Explain how all the internal events modeling updates performed to resolve internal event F&Os have an inconsequential impact on the RICT calculations contribution from FPRA.

Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty 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. The implementation of some of the complex FPRA methods often use nonconservative and over-simplified assumptions to apply the method to specific plant configurations. Historically, some of these issues were not always 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 fire risk profile of the plant.

The NRC staff evaluates the acceptability of the PRA for each new risk-informed application and, as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision making may vary with the relative weight given to the risk assessment element of the decisionmaking process. The calculated results of the PRA are used directly to calculate a RICT, which subsequently determines how long SSCs (both individual SSCs and multiple unrelated SSCs) controlled by technical specifications can remain inoperable. Therefore, the PRA results are given a very high weight in a TSTF-505 application, and the NRC is asking for information on the following issues that have been previously identified in previous TSTF-505 LARs as potentially key FPRA assumptions.

Reduced Transient Heat Release Rates (HRRs)

AUDIT QUESTION Q-24 The letter, Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires (ADAMS Accession No. ML12172A406), dated June 21, 2012, from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, discusses the key factors in justifying using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 (ADAMS Accession Nos. ML052580075 and ML052580118). Pages 11 and 12 of EC-RISK-1181, Revision 2 in the audit portal indicates that 60 kW and 145 kW maximum transient combustible fuel fires were assumed for small rooms where: storage is not allowed, is controlled, or is not practicable; maintenance is not required or is limited; and occupancy is limited.

Therefore:

Confirm whether the licensee used any reduced transient HRRs below the bounding 98 percent HRR of 317 kilowatts from NUREG/CR-6850. If yes, then EITHER:

Demonstrate that using reduced transient HRRs has an insignificant impact on this application.

OR:

Justify the use of the reduced HRRs, including:

Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR, including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance.

Also, include discussion of the personnel traffic that would be expected through each location.

The results of a review of records related to compliance with the transient combustible and hot work controls.

Treatment of Sensitive Electronics AUDIT QUESTION Q-25 FAQ 13-0004, Clarifications on Treatment of Sensitive Electronics (ADAMS Accession No. ML13322A085), provides supplemental guidance for applying the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850, Volume 2, for solid-state and sensitive electronics. The audit portal document EC-RISK-1181, Revision 2 (pages 14 and 16) states that FAQ 13-0004 provides guidance on modeling sensitive electronics but does not explain whether the guidance was applied, including the caveats on its use.

a.

Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents).

b.

If the treatment of sensitive electronics for the FPRA includes deviations from FAQ 13-0004, then:

EITHER:

Identify the deviations, and justify (e.g., through a sensitivity calculation) that the treatment of sensitive electronics has no consequential impact on the RICT calculations.

OR:

Identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09-A, prior to implementation of the RICT program.

Minimum Joint Human Error Probability AUDIT QUESTION Q-26 NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis GuidelinesFinal Report (ADAMS Accession No. ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in human reliability analyses (HRAs). NUREG-1921 refers to Table 2-1 of NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA) (ADAMS Accession No. ML051160213), which recommends that joint human error probability (HEP) values should not be below 1E-5. Table 4-4 of Electrical Power Research Institute (EPRI) 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, EPRI, Palo Alto, CA, 2010 (available on EPRIs Web site), provides a lower limiting value of 1E-6 for sequences with a very low level of dependence.

Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency.

The LAR does not provide this information and does not explain what minimum joint HEP value is currently assumed in the fire FPRA. Also, even if the assumed minimum joint HEP values are shown to have no impact on the current FPRA risk estimates, it is not clear to the NRC staff how the licensee will ensure that the impact remains minimal for future PRA model revisions.

Page 87 of the portal document EC-RISK-1185 indicates that the minimum joint HEP is 1E-06 (i.e., the same as the internal events minimum joint HEP). However, if one or more of the independent constituent actions within the JHEP has more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> available, the floor value applied is 5.0E-07. Therefore:

a.

Confirm and provide the minimum joint HEP value assumed in the FPRA.

b.

EITHER:

If the FPRA used a minimum joint HEP value of less than 1E-05, then demonstrate (e.g., through a sensitivity study) that the minimum joint HEP value(s) used have an inconsequential impact on the RICT application. If a sensitivity study is performed, then provide a description of and the quantitative results from the sensitivity study.

OR:

If the licensee cannot justify that the minimum joint HEP value has an inconsequential impact on the application, then:

Confirm that each FPRA joint HEP value below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline (i.e., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence). Provide an estimate of the number of these joint HEP values below 1.0E-5, discuss the range of values, and provide at least two different examples where this justification is applied.

If the licensee cannot justify joint HEP values used in the fire PRA below 1E-5, then identify appropriate RMAs for this key assumption, consistent with the treatment of key assumptions in NEI 06-09-A, prior to the implementation of the RICT program.

Well-Sealed Motor Control Center (MCC) Cabinets AUDIT QUESTION Q-27 Guidance in FAQ 08-0042, Fire Propagation from Electrical Cabinets, from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440 volts (V). With respect to Bin 15, as discussed in Chapter 6 of this supplement, the FAQ clarifies the meaning of robustly or well-sealed. Thus, for cabinets of 440 V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440 V and higher, the original guidance in Chapter 6 remains and requires that Bin 15 panels that house circuit voltages of 440 V or greater are counted because an arcing fault could compromise panel integrity (i.e., an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires). FPRA FAQ 14-0009, Treatment of Well-Sealed MCC Electrical Panels Greater than 440V (ADAMS Accession No. ML15119A176), provides the technique for evaluating fire damage from MCC cabinets having a voltage greater than 440 V. Propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440 V or greater. Therefore:

Describe how fire propagation outside of well-sealed MCC cabinets greater than 440 V is evaluated. If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, then provide justification for using this approach.

Transient Fire Influencing Factors AUDIT QUESTION Q-28 NUREG/CR-6850, Section 6, and FAQ 12-0064, Hot Work/Transient Fire Frequency Influence Factors (ADAMS Accession No. ML12346A488), describe the process for assigning influence factors for hot work and transient fires. The fire ignition frequency report (i.e., Report EC-RISK-1176, Revision 2) discusses apportionment of the transient ignition source frequency using the guidance and weighting values provided in FAQ 12-0064 with the exception that the very high influence factor, which has a weighting factor of 50, applicable to the Maintenance category was not used at all. It appears to the NRC staff that the very high influence factor may be applicable in some of the physical analysis units (PAUs) based on the guidance in FAQ 12-0064. Therefore:

Provide a sensitivity study or other justification demonstrating that assigning weighting factors of 50 per the guidance in FAQ 12-0064 has an inconsequential impact on the RICT calculations.

PRA Treatment of Fire Dependencies Between Units 1 and 2 AUDIT QUESTION Q-29 Many plants have Unit 1 and 2 adjoined and, thus, have common areas. For these plants, the risk contribution from fires originating in one unit must be addressed for impacts to the other unit given the physical proximity of the other unit and common areas.

Therefore, confirm whether Units 1 and 2 have common areas and shared systems, and if yes, then:

a. Explain how the risk contribution of fires originating in one unit is addressed for the other unit given the impacts from the physical proximity of equipment and cables in one unit to equipment and cables in the other unit. Include identification of locations where fire in one unit can affect components in the other unit.

b. Explain how the FPRA addresses contributions of fires in common areas that can impact components in both units, including the risk contribution of such scenarios. If any such scenarios are not addressed in the FPRA, then provide justification that their exclusion does not impact this application.

Dispositions of Fire PRA Model Assumptions and Sources of Uncertainty AUDIT QUESTION Q-30 LAR Enclosure 9, Table E9-2, states that the Human Reliability Analysis (HRA) for the fire PRA was performed using industry consensus modeling approaches but does not cite use of NRCs most current guidance on fire HRA, NUREG-1921. It is not clear if the licensee considered this guidance in its HRA. Therefore:

a. Confirm whether the licensee considered the guidance in NUREG-1921 to perform the HRA for the fire PRA. If it did, then describe any deviations from the guidance that could be characterized as potential key assumptions or sources of modeling uncertainty.

b. If in response to part (a) deviations from the guidance in NUREG-1921 could be characterized as assumptions or sources of modeling uncertainty, then justify that this modeling uncertainty has an inconsequential impact on the application (e.g., by performing a sensitivity study). If it cannot be determined that this modeling uncertainty has an inconsequential impact on the estimated RICTs, then identify programmatic changes to compensate for this uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty).

AUDIT QUESTION Q-31 The assessment of key assumptions and sources of uncertainty report (EC-RISK-0056) identifies treatment of unknown cable locations as a potential key of source of model uncertainty. In Table 2.3-1, Topic 40, the report states, It is not uncommon to not know specifically in a room where every cable is located. As a result, the FPRA assumes the cable is damaged for every fire until the cable is traced in detail. In Table 2.3-1, Topic 12, the report also states, Assumed cable routing was not applied as a part of the FPRA.

The uncertainty analysis concluded that this modeling assumption was not a key source of uncertainty because assumed cable routing was not used for cables with unknown locations; rather, these cables were assumed to fail in all fire scenarios. NRC notes that although this assumption is conservative in terms calculating total fire risk, it could have a nonconservative impact on the RICT calculations. If an SSC is part of a system not credited in the fire PRA, or if it is supported by a system that is assumed to always fail, then the increase in risk caused by taking that SSC out of service is masked. Neither the LAR nor the uncertainty analysis identified which components were assumed to be failed caused by the lack of cable routing information or what impact this assumption has on the RICT calculations. Therefore:

EITHER:

Identify the systems or components that are assumed to be always failed in the fire PRA (or are not included in the fire PRA) caused by a lack of known cable routing and justify that this assumption has an inconsequential impact on the RICT calculations.

OR:

If it cannot be determined that the cited assumption has an inconsequential impact on the estimated RICTs, then identify programmatic changes to compensate for this uncertainty and the basis for them (e.g., identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty).

Probabilistic Risk Analysis Licensing Branch C External Hazards Questions Determination of Seismic CDF Penalty A seismic PRA model is not available for Susquehanna, and the seismic hazard cannot be screened out for the RICT application. Therefore, the licensee provided the details of an approach for determining the seismic penalty in Section 3 of Enclosure 4 to the LAR. The LAR states that seismic CDF was estimated based on the current Susquehanna seismic hazard developed in response to the Near-Term Task Force (NTTF) Recommendation 2.1 and a plant-level high confidence of low probability of failure (HCLPF) capacity of 0.30g referenced to peak ground acceleration (PGA). NRC staff notes that the HCLPF capacity of 0.3g is higher than the HCLPF capacity of 0.21g cited for Susquehanna in Table B.2 and C-2 of NRC Generic Issue 199 (GI-199), Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, Safety/Risk Assessment Results, dated September 2, 2010 (ADAMS Accession No ML100270582). The LAR indicates that outliers were identified in the Susquehanna Individual Plant Examination for External Events (IPEEE) for SSCs with capacities lower than 0.3g that were reviewed and resolved in a supporting document to the LAR (Susquehanna Calculation EC-RISK-0045, High Confidence Low Probability Failure (HCLPF) Value for the Seismic Penalty Calculation - Seismic Margin Analysis for Risk Informed Applications, dated August 28, 2020). However, the justification for using a HCLFP of 0.3g to estimate a bounding seismic CDF despite plant SSCs with capacity of 0.21g is not provided in the LAR.

AUDIT QUESTION Q-32 Justify that using a plant-level HCLFP capacity of 0.3g is sufficient for estimating a bounding seismic CDF for this application given that there are plant SSCs with a HCLFP capacity of 0.21g. Include discussion of the cited outliers that were identified in the Susquehanna IPEEE and the resolutions of those outliers for this application.

AUDIT QUESTION Q-33 Confirm whether there are outliers that include SSCs with a lower HCLPF capacity than 0.21g and whose failure can impact seismic CDF. If yes, then justify that using the lower seismic capacity for those components has an inconsequential impact on the RICT calculations.

Determination of Seismic LERF Penalty LAR Enclosure 4, Section 3 states that the proposed bounding seismic LERF estimate is based on multiplying the estimated seismic CDF (i.e., 1.7E-05 per year) by an average seismic Conditional Large Early Release Probability (CLERP). The LAR states that the licensee does not have plant-specific information on seismic accident sequences or CLERPs and, therefore, it used information from plants of similar design (i.e., Limerick Generating Station (LGS) and LaSalle County Station (LCS)). Based on the seismic information from three sources for these two plants, the licensee determined the percent CDF contribution of each seismic accident scenario type to the total in LAR Table E4-3. Separately, the seismic CLERPs for seismic accident scenarios were estimated using the CLERP and other information from the Susquehanna internal events PRA. It appears that the licensee multiplied the percent contributions of the different accident types to seismic CDF by the estimated seismic scenario CLERPs to produce seismic sequence-weighted CLERPs for designated accident types.

However, the LAR does not present or discuss these calculations. Rather, the LAR states that based on the information in LAR Table E4-3, an estimate of seismic CLERP for Susquehanna could range from 3.3E-02 to 0.45 and that a CLERP of 0.24 is the average of the highest and lowest estimated seismic CLERPs. The LAR also states that for the purpose of calculating a seismic LERF penalty, the estimated CLERP was raised to 0.30.

It is not clear to NRC staff how the 3.3E-02 to 0.45 values were determined or why using the average estimated seismic CLERP from these determinations (which is adjusted slightly up) is sufficient for estimation of the seismic LERF. NRC staff notes the large difference presented in LAR Table E4-3 between the percent contributions of CDF scenarios that go directly to core damage for LGS and LCS. NRC staff also notes that the seismic studies for the LGS and LCS are not very recent (e.g., 30-40 years old) and, therefore, it is not clear whether using more current hazard or plant design information, or using Susquehanna-specific information, would produce different seismic scenario CDF contributions that would impact the application.

Therefore:

AUDIT QUESTION Q-34 Clarify how the percent seismic CDF contributions by sequence presented in the second, third, and fourth columns of LAR Table E4-3 and the seismic CLERP estimates for Susquehanna presented in the fifth column of the table were combined to determine the final sequence-weighted seismic CLERP values presented in the last row of the table (i.e., 0.36, 0.45, and 3.3E-02) for the three data sources.

AUDIT QUESTION Q-35 Justify why using the average of the large difference presented in LAR Table E4-3 in the percent contribution of seismic CDF scenarios that go directly to core damage for LGS and LCS is sufficient to support estimation of the LERF seismic penalty for Susquehanna, given that Susquehanna and Limerick use the GE Type 4 reactor while LCS uses the GE Type 5 reactor. Explain why the percent contribution of this scenario to the total seismic CDF for Susquehanna is not more like LGS (i.e., 33 or 43 percent) than it is like LCS (<1 percent).

AUDIT QUESTION Q-36 Justify that using seismic information about the percent contribution of seismic CDF to the total from older seismic studies for LGS and LCS provide a sufficient basis for estimating the seismic LERF penalty for this application. Discuss the review of more current hazard or plant design information with the potential to impact the seismic accident scenario CDF contributions to the total seismic CDF. Discuss site and plant differences that could impact the seismic accident scenario CDF contributions to the total seismic CDF.

External Flooding AUDIT QUESTION Q-37 Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A, states, [I]mpact 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 implementation, that external events that are not modeled in the PRA are not significant contributors to configuration risk. NRCs SE for NEI 06-09 states, [O]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.

LAR Enclosure 4 explains that the consequences of the reevaluated flood mechanisms, including local intense precipitation (LIP), were considered bounded by the plants current licensing design basis. The licensee explained that the reevaluated LIP would not result in flooding of any safety-related SSCs because of the presence of permanently installed watertight doors in the Reactor Buildings and Engineered Safeguard Service Water (ESSW) Pumphouse.

The LAR did not describe any RMAs 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.

Identify and justify the mechanism that will be used to ensure that the watertight doors will be closed during a flood event to prevent impact on risk significant equipment.

List of Audited Documents The U.S. Nuclear Regulatory Commission (NRC) staff audited, in a sampling manner, the following licensee documents during its review of the license amendment request for the Susquehanna Steam Electric Station (Susquehanna), Units 1 and 2. This enclosure does not include documents that are in NRCs Agencywide Documents Access and Management System (ADAMS).

Calculations and Reports DCP-95-9047, Elimination of Seismic Gaps Between Plant Panels, dated February 22, 1996 DCP-95-9048, Elimination of Seismic Gaps Between Plant Panels, dated March 27, 1996 EC-0113-1860, Handling of Transient Combustibles in the Wraparound Zones and restricted Areas (Red Zones) draft EC-0093-1023 NERPM-QA-0221-1, Revision 3, Turbine Missile Probability Analysis for Susquehanna Units 1 & 2, dated August 29, 2014 EC-Risk-0041, Revision 0, Containment Isolation and Containment Vent - PRA System Notebook, August 8, 2020 EC-Risk-0043, Revision 1, Attachment C Calculation, Technical Adequacy of the SSES PRA Models - Probabilistic Risk Analysis), dated October 7, 2020 EC-Risk-0045 Attachment C Calculation, Revision 0, High Confidence Low Probability of Failure (HCLPF) Value for the Seismic Penalty Calculation - Seismic Margin Analysis for Risk Informed Applications, dated August 28, 2020 EC-RISK-0046, Revision 1, Seismic CDF and LERF Estimate for TSTF-505 (RICT)

Program - Seismic Margin Analysis for Risk Informed Applications), dated March 17, 2021 EC-RISK-0046, Revision 2, Seismic CDF and LERF Estimate for TSTF-505 (RICT)

Program - Seismic Margin Analysis for Risk Informed Applications), dated December 9, 2021 o Excel Workbook, Susquehanna SCDF-SLERF Estimate_Rev 0.xls o Excel Workbook PA-B-NA-502_R0.xls Susquehanna Seismic Penalty Results for RICT Calculations (superseded) o Excel Workbook PA-B-NA-502_R2.xls Susquehanna Seismic Penalty Results for RICT Calculations EC-Risk-0047 Attachment C Calculation, Revision-0, External Hazards Assessment - SSES External Hazards Assessment for Risk Informed Applications, dated August 28, 2020 EC-Risk-0048 Attachment C Calculation, Revision 1, Fire PRA Interim Model OCT17R2F1 Quantification and Sensitivity - Fire Probabilistic Risk Analysis (FPRA),

dated September 22, 2020 EC-Risk-0055, Revision 1, Attachment C Calculation, Estimates for TSTF-505 (RICT)

Program LAR Submittal, dated March 17, 2021 EC-Risk-0056, Revision 0, Attachment C Calculation, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications, dated February 8, 2021 EC-Risk-1103, Revision 2, Attachment C Calculation, PSA-004.03 - SBLC System Notebook, dated June 25, 2015 EC-Risk-1104, Revision 1, Attachment C Calculation, PSA-004.02 - Manual Rod Insertion (MRI), Alternate Rod Insertion (ARI), and Recirculation Pump Trip (RPT and EOC-RPT) System Notebook, dated December 17, 2012 EC-Risk-1105, Revision 1, Attachment C Calculation, PSA-004.20 - Turbine Building Closed Cooling Water Notebook, dated January 8, 2013 EC-Risk-1106, Revision 3, Attachment C Calculation, PSA-004.6 - RCIC PRA System Notebook, dated November 10, 2017 EC-Risk-1108, Revision 1, Attachment C Calculation, PSA-004.04 - CRS Spray System Notebook, dated December 18, 2012 EC-Risk-1109, Revision 3, Attachment C Calculation, PSA-004.18 - Emergency SW Notebook, dated January 8, 2013 EC-Risk-1110, Revision 2, Attachment C Calculation, PSA-004.17 - RHRSW System Notebook, dated December 13, 2012 EC-Risk-1111, Revision 2, Attachment C Calculation, PSA-005.5 - HPCI System Notebook, dated February 13, 2012 EC-Risk-1112, Revision 2, Attachment C Calculation, PSA-004.23 - Instrument Air and Service Air System Notebook, dated December 12, 2012 EC-Risk-1113, Revision 2, Attachment C Calculation, PSA-004.22 - Service Water System Notebook, dated December 13, 2012 EC-Risk-1114, Revision 2, Attachment C Calculation, PSA-004.09 - Core Spray System Notebook, dated January 8, 2013 EC-Risk-1115, Revision 3, Attachment C Calculation, PSA-004.12 - ADS & Main Steam Isolation Valve PRA System Notebook, dated November 10, 2017 EC-Risk-1116, Revision 1, Attachment C Calculation, PSA-004.08 - Condensate System Notebook, dated January 8, 2013 EC-Risk-1117, Revision 1, Attachment C Calculation, PSA-004.27 - Reactor Vessel Instrumentation System Notebook, dated January 8, 2013 EC-Risk-1118, Revision 1, Attachment C Calculation, PSA-004.25 - Electrical Distribution PRA System Notebook, dated March 28, 2008 EC-Risk-1119, Revision 1, Attachment C Calculation, PSA-004.21 - Reactor Building Closed Cooling Water Notebook, dated December 13, 2012 EC-Risk-1120, Revision 2, PSA-004.11 - Fire Protection Water PRA System Notebook, November 10, 2017 EC-Risk-1120, Revision 2, Attachment C Calculation, PSA-004.11 - Fire Protection Water PRA System Notebook, dated November 10, 2017 EC-Risk-1131, Revision 3, Attachment C Calculation, PSA-004.16 - Residual Heat Removal (RHR) System Notebook, dated January 8, 2013 EC-Risk-1139, Revision 7, Attachment C Calculation, Susquehanna PRA Model Event Tree Notebook and Success Criteria Post-EPU Level 2), dated October 26, 2020 EC-Risk-1144, Revision 2, Attachment C Calculation, PSA-004.24 - Containment Instrument Gas PRA System Notebook, dated November 10, 2017 EC-Risk-1153, Revision 0, Attachment C Calculation, PSA-004.7 - Feedwater System Notebook, dated December 17, 2012 EC-Risk-1156, Revision 0, Attachment C Calculation, PSA-004.10 - HVAC System Notebook, dated December 17, 2012 EC-Risk-1157, Revision 0, Attachment C Calculation, PSA-004.18 - DC PRA System Notebook, dated November 10, 2017 EC-Risk-1158, Revision 1, EDG and AC PRA System Notebook, November 10, 2017 EC-Risk-1158, Revision 1, Attachment C Calculation, PSA-004.19 - EDG and AC PRA System Notebook, dated November 10, 2017 EC-Risk-1160, Revision 0, Attachment C Calculation, PSA-004.13 - CST and RWST System Notebook, dated December 17, 2012 EC-Risk-1172, Revision 1, Attachment C Calculation, Plant Partitioning (PP) and Qualitative Screening (QLS) - Fire Probabilistic Risk Analysis (FPRA),

September 15, 2020 EC-Risk-1173, Revision 2, Attachment C Calculation, Equipment Selection (ES) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1174, Revision 2, Attachment C Calculation, Fire PRA Cable Selection (CS) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1175, Revision 1, Attachment C Calculation, Plant Response Model (PRM) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1176, Revision 2, Attachment C Calculation, Fire Ignition Frequency (IGN) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1177, Revision 2, Attachment C Calculation, Fire Scenario Selection (FSS) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1178, Revision 1, Attachment C Calculation, Structural Steel Analysis (SS) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1179, Revision 1, Attachment C Calculation, Multi-Compartment Analysis (MCA) - Fire Probabilistic Risk Analysis (FPRA), dated September 15, 2020 EC-Risk-1180, Revision 1, Attachment C Calculation, Walkdowns - Fire Probabilistic Risk Analysis (FPRA), dated September 16, 2020 EC-Risk-1181, Revision 2, Attachment C Calculation, Fire Modeling Treatment (FMT) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1182, Revision 2, Attachment C Calculation, Main Control Room Abandonment Analysis (MCRAB) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1184, Revision 1, Attachment C Calculation, Circuit Failure Mode Likelihood Analysis (CF) - Fire Probabilistic Risk Analysis (FPRA), dated September 16, 2020 EC-Risk-1185, Revision 2, Attachment C Calculation, Human Reliability Analysis (HRA) - Fire Probabilistic Risk Analysis (FPRA), dated September 21, 2020 EC-Risk-1186, Revision 1, Attachment C Calculation, Seismic Fire Interactions (SF),

dated September 21, 2020 EC-Risk-1187, Revision 1, Attachment C Calculation, Fire Risk Quantification (FQ) and Uncertainty and Sensitivity Analysis (UNC), dated September 16, 2020 NQPA-B-NA-009, SSES Full Power Internal Events Probabilistic Risk Assessment - 2020 F&O Roadmap, dated September 4, 2020 NQPA-B-NA-012, SSES Full Power Internal Events Probabilistic Risk Assessment, November 1, 2021 One-Line Diagrams E-1 125 and 250 VDC for Units 1 and 2 E-1 Switchgear to 4.16 kV for Units 1 and 2 E-5 4.16 kV for Units 1 and 2 E-8 480 VAC for Units 1 and 2 Miscellaneous Documents Presentation Slides on Susquehanna Steam Electric Station EPRI Phoenix Risk Monitor

- Software Features that Facilitate Risk Management Implementation of Risk-Informed Extended Completion Times - RITSTF Initiative 4b - NRC TSTF-505 Audit Demo -

November 10, 2021 Unit 1 ESS Bus Load List from ON-4KV-101, Loss of 4KV Bus, Revision 8, Attachment D Unit 2 ESS Bus Load List from ON-4KV-201, Loss of 4KV Bus, Revision 6, Attachment D Piping and Instrumentation Diagrams M-111 Emergency Service Water for Units 1 and 2 M-112 RHR Service Water for Units 1 and 2 M-141_Nucelar Boiler for Units 1 and 2 M-148_Standby Liquid Control for Units 1 and 2 M-149_RCIC for Units 1 and 2 M-151_RHR for Units 1 and 2 M-152_Core Spray for Units 1 and 2 M-155_HPCI for Units 1 and 2 Procedures NDAP-QA-0340, Revision 36, Protected Equipment List NDAP-QA-0440, Revision 15, Control of Transient Combustible/Hazardous Materials NDAP-QA-1201, Configuration Management Process and Programs, dated March 17, 2021 NFP-QA-201, Revision 4, Internal Events at Power PRA Model Update and Configuration Control Process

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 -

SUMMARY

OF REGULATORY AUDIT IN SUPPORT OF RISK-INFORMED COMPLETION TIMES IN TECHNICAL SPECIFICATIONS LICENSE AMENDMENT REQUEST (EPID L-2021-LLA-0062) DATED MARCH 31, 2022 DISTRIBUTION:

PUBLIC PM File Copy RidsRgn1MailCenter Resource RidsACRS_MailCTR Resource RidsNrrDorlLpl1 Resource RidsNrrPMSusquehanna Resource RidsNrrLAKZeleznock Resource RidsNrrDexEeeb Resource RidsNrrDexEicb Resource RidsNrrDexEmib Resource RidsNrrDraApla Resource RidsNrrDraAplb Resource RidsNrrDraAplc Resource RidsNrrDssScpb Resource RidsNrrDssSnsb Resource RidsNrrDssStsb Resource NKhan, NRR KNguyen, NRR MLi, NRR JZhao, NRR KHsu, NRR GBedi, NRR MPatterson, NRR THilsmeier, NRR NIqbal, NRR CMoulton, NRR SPark, NRR KTetter, NRR HWagage, NRR NChien, NRR ASallman, NRR KWest, NRR GMiller, NRR ADAMS Package Accession No. ML22070B130 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EEEB NAME AKlett KZeleznock NKhan DATE 03/11/2022 03/16/2022 03/10/2022 OFFICE NRR/DEX/EEEB NRR/DEX/EEEB/BC NRR/DEX/EICB NAME KNguyen WMorton MLi DATE 03/09/2022 03/17/2022 03/09/2022 OFFICE NRR/DEX/EICB NRR/DEX/EICB/BC NRR/DRA/APLA NAME JZhao MWaters MPatterson DATE 03/10/2022 03/14/2022 03/10/2022 OFFICE NRR/DRA/APLA NRR/DRA/APLA/BC NRR/DRA/APLB NAME THilsmeier RPascarelli NIqbal DATE 03/10/2022 03/24/2022 03/10/2022 OFFICE NRR/DRA/APLB NRR/DRA/APLB/BC NRR/DRA/APLC NAME CMoulton JWhitman Sunwoo Park DATE 03/10/2022 03/22/2022 03/09/2022 OFFICE NRR/DRA/APLC NRR/DRA/APLC/BC NRR/DSS/SCPB NAME KTetter SRosenberg (SVasavada for)

HWagage DATE 03/10/2022 03/11/2022 03/10/2022 OFFICE NRR/DSS/SCPB NRR/DSS/SCPB/BC NRR/DSS/STSB NAME NChien BWittick KWest DATE 03/17/2022 03/16/2022 03/09/2022 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME VCusumano JDanna AKlett DATE 03/24/2022 03/31/2022 03/31/2022 OFFICIAL RECORD COPY