ML24110A049

Audit Report Related to the TSTF-505 and 10 CFR 50.59 Amendments

ML24110A049
Person / Time
Site:	Quad Cities
Issue date:	05/28/2024
From:	Robert Kuntz Plant Licensing Branch III
To:	Rhoades D Constellation Energy Generation
Kuntz R
References
EPID L-2023-LLA-0084, EPID L-2023-LLA-0085
Download: ML24110A049 (1)

Text

May 28, 2024 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2, -

REGULATORY AUDIT REPORT TO SUPPORT THE REVIEW OF THE AMENDMENTS TO ADOPT TSTF-505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B, AND 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS, (EPID L-2023-LLA-0084 AND EPID L-2023-LLA-0085)

Dear David P. Rhoades:

By letters dated June 8, 2023, Constellation Energy Generation, LLC (the licensee) submitted license amendment requests (LARs) to amend the licenses for Quad Cities Nuclear Power Station, Units 1 and 2, Renewed Facility Operating License Nos. DPR-29 and DPR-30, respectively. The proposed LARs (Agencywide Documents and Access Management System (ADAMS) Accession Nos. ML23159A249 and ML23159A253) would adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-informed Extended Completion Times, RITSTF [Risk-informed Technical Specifications Task Force]

Initiative 4b and adopt the provisions of Title 10 of the Code of Federal Regulations (10 CFR),

section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

Due to the complex nature of this material, the U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit for understanding under the Office of Nuclear Reactor Regulation guidance LIC-111, Regulatory Audits (ML19226A274). The audit plan was provided on November 20, 2023 (ML23319A334).

The audit was conducted from November 20, 2023, through March 31, 2024. The audit was conducted virtually using an online portal and online discussions. The purpose of the audit was to gain understanding, to verify information, and to identify information that will require docketing to support the proposed licensing actions. The enclosure to this letter provides a report of the NRC staffs audit.

D. Rhodes If you have any questions, please contact me by telephone at (301) 415-3733 or by e-mail at Robert.Kuntz@nrc.gov.

Sincerely,

/RA/

Robert Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosure:

1. Audit Report

2. Audit Questions cc: Listserv

REGULATORY AUDIT REPORT TO SUPPORT THE REVIEW OF LICENSE AMENDMENT REQUESTS TO ADOPT RISK INFORMED COMPLETION TIMES - TSTF-505 AND 10 CFR 50.69 RISK-INFORMED CATEGORIZATION OF SYSTEMS, STRUCTUES, AND COMPONENTS QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265

1.0 BACKGROUND

By letters dated June 8, 2023, Constellation Energy Generation, LLC (the licensee) submitted license amendment requests (LARs) for Quad Cities Nuclear Power Station (Quad Cities), Units 1 and 2. The proposed LARs (Agencywide Documents and Access Management System (ADAMS) Accession No. ML23159A249 and ML23159A253) requested to modify the Quad Cities technical specifications (TSs) to permit the use of risk-informed completion times (RICTs) in accordance with Technical Specifications Task Force (TSTF)-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ML18183A493), and proposed the addition of a license condition that allows implementation of the provisions in Title 10 of the Code of Federal Regulations (10 CFR), section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors.

The U.S. Nuclear Regulatory Commission (NRC) staff determined it would be efficient to conduct an audit in order to enable the staff to examine and evaluate information with the intent to gain understanding, verify information, and/or identify information that will require docketing to support the basis of a licensing or regulatory decision. The NRC staff performed an audit from November 20, 2023, through March 31, 2024, including formal audit discussion (breakout) topic sessions that were conducted with the licensee the week of January 22, 2024, and on March 19, 2024.

The NRC staff performed and completed its audit in accordance with NRR Office Instruction LIC-111, Revision 1, Regulatory Audits (ML19226A274).

2.0 AUDIT TEAM Section 5.0 of the NRC audit plan issued on November 20, 2023 (ML23319A334), provides the list of audit team members, in addition to those listed in the plan the following NRC staff participated in the audit:

Name Review Area Hosung Ahn probabilistic flood hazard analysis Amy DAgostino human factors 3.0 AUDIT

SUMMARY

The NRC staff reviewed the documents listed in section 4.0 and conducted discussions with the licensee staff. The NRC staff interaction with the licensee staff included discussion of the audit questions included in Enclosure 2.

4.0 DOCUMENTS REVIEWED The following documents were reviewed by the NRC staff during the audit:

Record Type Document Number Title Revision or Date PRA Peer Review 004N2062-R0-QC Quad Cities Nuclear Generating Station Units 1 and 2 PRA Peer Review Report 4/3/2017 PRA Peer Review 032366-RPT-001 Rev 1 Quad Cities PRA Finding Level Fact and Observation Independent Assessment Revision 1 PRA Peer Review 032466-RPT-007 Rev 0 Quad Cities PRA Finding Level Fact and Observation Independent Assessment Revision 0 PRA Peer Review 032466-RPT-008 Rev 0 Quad Cities Unit 1 PRA Focused-Scope Peer Review Revision 0 PRA Peer Review QC-PFRA-PR-Merged-Report-Final-09-25-2013 Quad Cities Generation Station Unit 1 Fire PRA Peer Review Report September 2013 PRA Peer Review 032466RPT-01-Rev 0 Final Quad Cities PRA Focused-Scope Peer Review Revision 0 PRA Peer Review QC-MISC-048 (HW and tornados)

Quad Cities PRA Finding Level Fact and Observation Independent Assessment; High Winds and Tornados Revision 0 May 2021 Record Type Document Number Title Revision or Date RICT Calculation QC-MISC-049-R0 RICT Estimates for TSTF-505 (RICT) Program LAR Submittal Revision 0 RICT Calculation ER-AA-600-1053, Rev 000 Calculation of RMAT and RICT for Risk Informed Completion Time Program Revision 0 FLEX Program LN-FLEX.2 Rev 006 FLEX EQUIPMENT Revision 6 FLEX Program EC 404409, 20160420 INTEGRATED REVIEW OF FLEX ACTIONS IAW NEI VALIDATION PROCESS FUKUSHIMA April 2016 PRA Program OP-AA-108-117, REV 007 Protected Equipment Program Revision 7 PRA Program OP-AA-108-118, REV 003 Risk Informed Completion Time Revision 3 PRA Program WC-AA-101-1006, REV 004 On-line Risk Management and Assessment Revision 4 PRA Program QC-MISC-047 R0 Assessment of Key Assumptions and Sources of Uncertainty for the Quad Cities Nuclear Power Station April 2023 PRA Program QC-MISC-039 Rev 0 Final External Hazards Assessment for Quad Cities Nuclear Power Plant Revision 0 PRA Program OP-AA-201-012-1001, REV 005 OPERATIONS ON-LINE FIRE RISK MANAGEMENT Revision 5 PRA Program OP-QC-201-012-1001, REV 009 QUAD CITIES ON-LINE FIRE RISK MANAGEMENT Revision 9 PRA Program Exelon Report No. SL-012196 Seismic Hazard and Screening Report March 2014 Training and Reference Material ER-AA-600-1015, REV 020 FPIE PRA Model Update Revision 20 Training and Reference Material ER-AA-600-1016, REV 014 Configuration Risk Model Update Revision 14 Training and Reference Material ER-AA-600-1023, REV 011 Paragon Model Capability Revision 11 Training and Reference Material ER-AA-600-1042, REV 013 On-line Risk Management Revision 13 Record Type Document Number Title Revision or Date Training and Reference Material ER-AA-600-1052, REV 001 Risk Management Support of RICT Revision 1 Training and Reference Material ER-AA-600-1061, REV 007 Fire PRA Model Update and Control Revision 7 Training and Reference Material ER-AA-600-1012, REV 014 Risk Management Documentation Revision 14 Training and Reference Material ER-AA-600-1015, REV 020 FPIE PRA Model Update Revision 20 Training and Reference Material ER-AA-600-1061, REV 007 Fire Model Update and Control Revision 7 Technical Specification Bases TS B 3.8.4 Technical Specification Bases 3.8.4, DC Sources-Operating Revision 0 Technical Specification Bases TS B 3.8.7 Technical Specification Bases 3.8.7, Distribution System - Operating Revision 52 Circuit Diagram 4E-1303 KEY DIAGRAM 4160V SWITCHGEAR 11 12 13 AND 14 (CRITICAL CONTROL ROOM DRAWING) 1/26/99 Circuit Diagram 4E-1303A KEY DIAGRAM 4160 VOLT SWITCHGEAR 13 (CRITICAL CONTROL ROOM DRAWING) 1/26/99 Circuit Diagram 4E-1304 KEY DIAGRAM 4160V SWITCHGEARS 13-1 AND 14-1 (CRITICAL CONTROL ROOM DRAWING) 10/10/96 Circuit Diagram 4E-1304A KEY DIAGRAM 4160 VOLT SWITCHGEAR BUSSES 13-1 AND 14-1 CRITICAL CONTROL ROOM DRAWING 10/10/96 Circuit Diagram 4E-1306 KEY DIAGRAM REACTOR BUILDING 480V SW GROUPS 18 AND 19 (CRITICAL CONTROL ROOM DRAWING) 9/5/01 Record Type Document Number Title Revision or Date Circuit Diagram 4E-1328 SINGLE LINE DIAG-EMERGENCY POWER SYSTEM UNIT 1 AND2 (CRITICAL CONTROL ROOM DRAWING) 9/5/01 Circuit Diagram 4E-2303 KEY DIAGRAM 4160V SWITCHGEARS 21, 22, 23 AND 24 (CRITICAL CONTROL ROOM DRAWING) 11/18/98 Circuit Diagram 4E-2303A KEY DIAGRAM 4160V SWITCHGEAR 24 (CRITICAL CONTROL ROOM DRAWING) 11/18/98 Circuit Diagram 4E-2304 KEY DIAGRAM 4160V SWITCHGEARS 23-1 AND 24-1 (CRITICAL CONTROL ROOM DRAWING) 5/9/97 Circuit Diagram 4E-2306-1 KEY DIAGRAM REACTOR BLDG 480V SWGR 28 (CRITICAL CONTROL ROOM DRAWING) 10/30/06 Circuit Diagram 4E-2306-2 KEY DIAGRAM REACTOR BLDG 480V SW GRPS 28, 29 (CRITICAL CONTROL ROOM DRAWING) 10/30/06 Circuit Diagram 4E-2312 KEY DIAG RX BLDG 480V MCC 28-1B, 29-4, 29-6_28,29-5 CRIB HOUSE 480V MCC 26-2 (CRIT CR DRAW) 11/1/04 Circuit Diagram 4E-6840 KEY DIAGRAM MCC 18-4 CONTROL ROOM HVAC UPGRADE (CRITICAL CONTROL ROOM DRAWING) 06/19/01 Circuit Diagram 4E-1317 KEY DIAGRAM 250V DC MOTORS CONTROL CENTERS 9/30/15 Circuit Diagram 4E-1318 KEY DIAGRAM 125V DC TURBINE BUILDING RESERVE BUS DISTRIBUTION PANEL AND 125V DC REACTOR BLDG DISTRIB 8/2/93 Record Type Document Number Title Revision or Date Circuit Diagram 4E-1318A KEY DIAGRAM TURBINE BUILDING 125V DC MAIN BUS DISTRIBUTION PANEL 4/22/98 Circuit Diagram 4E-1318B OVERALL KEY DIAGRAM 125V DC DISTRIBUTION CENTERS 10/17/00 Circuit Diagram 4E-2317 KEY DIAGRAM 250V DC MOTOR CONTROL CENTERS 4/28/97 Circuit Diagram 4E-2318 KEY DIAGRAM 125V DC DISTRIBUTION CENTER 8/2/93 Circuit Diagram 4E-2318A KEY DIAGRAM TURBINE BUILDING 125V DC MAIN BUS DISTRIBUTION PANEL 11/7/94 Circuit Diagram 4E-2318B OVERALL KEY DIAGRAM 125V DC DISTRIBUTION CENTERS 10/17/2000 System Diagram 0000392256 REV 002 Hardened Containment Vent System (Non-Outage Portion) 10/13/2017 System Diagram 0000392257 REV 002 Hardened Containment Vent System (Outage Portion) 4/10/2017 System Diagram 0000400666 REV 000 Hardened Containment Vent System 6/6/2017 PRA Notebook QC-PSA-004 HRA REV 9 Quad Cities Probabilistic Risk Assessment Human Reliability Notebook June 2019 PRA Notebook QC-PSA-005.27 PT EQP Quad Cities Probabilistic Risk Assessment Portable Equipment System Notebook Revision 1 June 2019 PRA Notebook QC-PSA-013 Quad Cities PRA Summary Document Notebook Revision 8 PRA Notebook QC-PSA-014 Quad Cities PRA Quantification Revision 5 PRA Notebook QC-PRA-021.62 QC FPRA Uncertainty and Sensitivity Notebook Revision 2 PRA Notebook QC-PSA-010 COMP DATA VOL 1 REV 8 Component Data Notebook Volume 1 2018 PRA Update PRA Notebook QC-PSA-010 COMP DATA VOL 2 REV 8 Component Data Notebook Volume 2 2018 PRA Update Record Type Document Number Title Revision or Date PRA Notebook QC-PSA-002 EVENT TREES REV 3 Event Tree Notebook 2018 PRA Update PRA Notebook QC-PSA-003 SUCCESS CRITERIA REV 6 Success Criteria Notebook 2018 PRA Update PRA Notebook QC-PSA-004 HRA REV 9 Human Reliability Analysis 2018 PRA Update PRA Notebook QC-PSA-005.01 RPS REV 5 Reactor Protection System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.02 RPT REV 4 Recirculation Pump Trip and Alternate Rod Insertion Systems Notebook 2018 PRA Update PRA Notebook QC-PSA-005.03 SBLC REV 6 Standby Liquid Control System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.04 CRD REV 5 Control Rod Drive Hydraulic System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.05 FW REV 7 Feedwater and Condensate System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.06 HPCI REV 7 High Pressure Coolant Injection System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.07 RCIC REV 7 Reactor Core Isolation Cooling System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.08 SSMP REV 6 Safe Shutdown Makeup Pump Notebook 2018 PRA Update PRA Notebook QC-PSA-005.09 ADS REV 3 Reactor Pressure Control and Automatic Depressurization System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.10 IA REV 6 Instrument Air System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.11 RHR REV 6 Residual Heat Removal System and Residual Heat Removal Service Water System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.12 FP REV 5 Fire Protection System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.13 CS REV 6 Core Spray System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.14 SW REV 8 Service Water System notebook 2018 PRA Update PRA Notebook QC-PSA-005.15 MC REV 3 Main Condenser and Man Steam System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.16 TDV REV 4 Torus-Drywell Vent System Notebook 2018 PRA Update Record Type Document Number Title Revision or Date PRA Notebook QC-PSA-005.17 TBCCW REV 6 Turbine Building Closed Cooling Water System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.18 RBCCW REV 4 Reactor Building Closed Cooling Water System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.19 VS REV 5 Vapor Suppression System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.20 EP REV 7 Electric Power System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.22 CAS REV 6 Common Actuation System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.23 PCIS REV 5 Primary Containment Isolation System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.24 OSWS REV 4 On-Site Water Sources System Notebook 2018 PRA Update PRA Notebook QC-PSA-005.25 DRYWELL COOLER REV 3 Drywell Cooler System Notebook 2018 PRA Update PRA Notebook QC-PSA-006 DEPENDENCY NB REV 7 Dependency Notebook 2018 PRA Update PRA Notebook QC-PRA-021.61 R2 Fire Probabilistic Risk Analysis Summary and Quantification Notebook Revision 2 Procedure QCOA 0010-16 REV 29 Flood Emergency Procedure Revision 29 Procedure QCOA 0010-14 REV 12 Lock and Dam #14 Failure Revision 12 Procedure QCOA 0010-17 REV 9 Toxic Gas/Chemical Release from Nearby Chemical Facilities Revision 9 Procedure QCOP 0010-01 REV 94 Winterizing Checklist Revision 94 Procedure QCOP 0010-10 REV 26 Required Hot Weather Inspections Revision 26 Procedure QCOA 0010-22 REV 17 Local Intense Precipitation Response Procedure Revision 17 Procedure QDC-0000-S-2089, REV 2 Evaluation of Flood Boundary Structures for Local Intense Precipitation January 2018 Procedure QUAD_CITIES_PFHA_REPORT_12-07-2021 Probabilistic Flood Hazard Assessment Report for the Mississippi River December 2021 Procedure EC 636912 REV0 Update to Station External Flood Response in Support of Risk Reductions Revision 0 Record Type Document Number Title Revision or Date Procedure DRAFT EC 636914 Draft Advanced Work Authorization Form Flood Boundary Revision 0 5.0 AUDIT CONCLUSION The purpose of the audit was to gain understanding, to verify information, and to identify information that will require docketing to support the proposed licensing action. As a result of the audit, the NRC staff issued a request for additional information on March 6, 2024 (ML24066A153), and April 10, 2024 (ML24102A242).

Date:

AUDIT QUESTIONS of the audit plan issued on November 20, 2023, included audit questions. In addition to the questions in the audit plan, the following questions were also discussed during the audit.

APLA Question 01 - Digital Instrumentation and Control (I&C) Modeling Concerning the quality of the probabilistic risk assessment (PRA) model, Nuclear Energy Institute (NEI) 06-09-A, Risk-Informed Technical Specifications Initiative 4b Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A (ML12286A322), states that Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (ML17317A256), and RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3 (ML20238B871), 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 technical specification (TS) change and the role the PRA plays in justifying the change.

Regarding digital instrumentation and control (I&C), the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist, such as 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. Also, although reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT and 10 CFR 50.69 programs.

Table E9-1 of Enclosure 9 to the RICT (TSTF-505) LAR and the table in Attachment 6 of the 10 CFR 50.69 LAR identifies the digital feedwater control (DFWC) failure probabilities as a potential key uncertainty and performed a sensitivity of increasing the likelihood that the DFWC would result in overfilling the reactor pressure vessel (RPV). The NRC staff notes that another failure mode consideration is loss of feedwater to the RPV. It is unclear to the NRC staff if the sensitivity study failure mode is bounding for this uncertainty. Therefore, address the following:

a) Provide justification that the DFWC failure mode addressed in the LAR sensitivity is the bounding failure mode of this control system.

b) If another failure mode is determined to be bounding, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations and 10 CFR 50.69 categorization.

Clarify whether digital I&C systems, other than DFWC, are credited in the PRA models that will be used in the RICT and 50.69 programs.

c) If other digital I&C systems are credited in the PRA models and will be used in the RICT or 10 CFR 50.69 programs, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations and 10 CFR 50.69 categorizations.

Alternatively, for RICT, if a justification is not provided identify which limiting condition for operations (LCOs) are determined to be impacted by digital I&C systems modeling for which risk management actions (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.

APLA Question 02 - Open Phase Condition Section C.1.4 of RG 1.200 states the base (e.g., Model of Record) PRA 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 necessitate changes to the base PRA model.

In response to the January 30, 2012, Open Phase Condition (OPC) event at the Byron Generating Station, 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). As per 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. From the revised voluntary initiative 4 and resulting industry guidance in NEI 19-02 5 on estimating OPC and OPIS risk, it is understood that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design.

In response to the NRC audit portal request regarding OPC and OPIS, the licensee stated that neither OPC nor OPIS is modeled in any Quad Cities PRA model. Considering these observations, provide the following information:

a) For Quad Cities, discuss the evaluation of the risk impact associated with OPC events including the likelihood of OPC initiating plant trips and the impact of those trips on PRA-modeled structure, system, and components (SSCs). Also, explain whether an OPIS has been installed and if it has been installed, then discuss its functionality and any operator actions needed to operate the system or needed in response to the system.

b) Provide justification that the exclusion of this failure mode and mitigating system does not impact RICT calculations or 10 CFR 50.69 categorizations.

c) As an alternative to Part (b), propose a mechanism to ensure that OPC-related scenarios are incorporated into the application PRA models prior to implementing the RICT or 10 CFR 50.69 programs.

1 U.S. NRC Bulletin 2012-01, Design Vulnerability in Electric Power System (ML12074A115).

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 (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 (ML17068A297).

4 Doug True to Ho Nieh, Ltr re: Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ML19163A176).

5 NEI 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, Revision 0, April 2019 (ML19122A321).

APLA Question 03 - PRA Model Uncertainty Analysis Results The NRC staff safety evaluation (SE) to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and sources of uncertainty and to assess/disposition each as to their impact on the RMTS application. LAR Enclosure 9, Table E9-1, identifies the key assumptions and sources of uncertainty for the internal events and fire PRA models and provides dispositions for each source of uncertainty for this TSTF-505 application. NRC staff reviewed these dispositions and is unclear that some uncertainty dispositions fully addressed the potential impact to RICT calculations. Therefore, address the following:

LAR Enclosure 9, Table E9-1, regarding core cooling success following containment failure, states that the sensitivity study demonstrates that all risk metrics are sensitive to this uncertainty.

NRC review of Section 8.1 of the Assessment of Key Assumptions and Sources of Uncertainty Notebook demonstrates impacts on core damage frequency (CDF) and large early release frequency (LERF) ranging from 33 to 81 percent. It appears that this source of uncertainty could plausibly impact this application. However, the LAR reasons that the increased factor value is not considered credible. The NRC staff notes that it requires insights from credible sensitivity studies for its review process and determination. It is unclear to the NRC staff that this assumption has no impact on the RICT or 10 CFR 50.69 programs. Therefore, address the following:

a) Clarify what increased factor value should be used for this sensitivity. Include in this discussion justification that the selected increased factor value appropriately bounds the increase in risk associated with this uncertainty.

b) Based on the response to part (a) above, provide justification that the uncertainty associated with core cooling success following containment failure does not significantly impact any RICT calculation or 10 CFR 50.69 categorizations.

c) Alternatively, to part (b) above, for RICT, provide how this source of uncertainty, such as additional RMAs, would be addressed in the RICT program.

APLA Question 04 - Consideration of Shared Systems in RICT Calculations RG 1.200, Revision 3, states, in part: The base PRA serves as the foundational representation of the as-built and as-operated plant necessary to support an application.

The LAR does mention the existence of interconnected auxiliary systems between units. The NRC staff notes that for some of these systems, it appears the sharing of a system is not consistent between units. Enclosure 8 to the LAR states that the Real Time Risk (RTR) model can represent the availability of these shared components. However, it is unclear to the NRC staff if all operational aspects, such as alternate alignments, were excluded from the PRA models. In addition, multi-unit events (e.g., loss of offsite power and seismic events), credit for a shared system may be limited to one unit.

Clarify what systems are shared, how they are shared, and whether they can support the other unit in an accident. Explain how the shared systems are credited for each unit in the PRA models. This discussion should also address the following:

a) Explain whether shared systems are credited in the internal events, including flood and fire PRA models for both units and, if so, identify those systems.

b) If shared systems are credited in the RTR model that supports the RICT calculations, then explain how the shared system is modelled for each unit in a dual unit event demonstrating that shared systems are not over-credited in the PRA models.

c) If a shared system is credited in the RTR model that supports the RICT calculations and the impact of events that can create a concurrent demand for the system shared by both units is not addressed in the PRA models, then justify that this exclusion has an inconsequential impact the RICT calculations.

APLA Question 05 - Impact of Seasonal Variations The Tier 3 assessment in RG 1.177, An Approach for Plant-specific, Risk-informed Decision-making: Technical Specifications, Revision 2 (ML20164A034), 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. NEI 06-09-A and its associated NRC SE (ML071200238) states 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.

of the LAR mentions modifications related to the impact of outside temperatures for one SSC. However, it does not appear to state if other modeling adjustments are needed to account for seasonal and time of cycle dependencies and what kind of adjustments will be made.

Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:

a) Explain how the RICT calculations address changes in PRA data points, basic events, and SSC operability constraints as a result of extreme weather conditions, seasonal variations, other environmental factors, or time of cycle. Also, explain how these adjustments are made in the configuration risk management program (CRMP) model and how this approach is consistent with the guidance in NEI 06-09-A and its associated NRC final SE.

b) Describe the criteria used to determine when PRA adjustments due to extreme weather conditions, seasonal variations, other environmental factors, or time of cycle variations need to be made in the CRMP model and what mechanism initiates these changes.

APLA Question 06 - Performance Monitoring The NRC SE for NEI 06-09-A, states, in part: The impact of the proposed change should be monitored using performance measurement strategies. NEI 06-09-A considers the use of NUMARC 93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (ML18120A069), as endorsed by RG 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4 (ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.

In addition, the NEI 06-09-A methodology satisfies the five key safety principles specified in RG 1.177, Revision 2, relative to the risk impact due to the application of a RICT. Moreover, NRC staff position C.3.2 provided in RG 1.177, Revision 2, for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period. It is unclear how the licensees RICT program captures performance monitoring for the SSCs within the scope of the RMTS program. Therefore:

a) Confirm that the Quad Cities Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in NUMARC 93-01, as endorsed by RG 1.160.

b) Alternatively, describe the approach or method used by Quad Cities for SSC performance monitoring, as described in NRC staff position C.3.2 of RG 1.177, Revision 2, 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-A.

APLA Audit Question 07 - In-Scope LCOs and Corresponding PRA Modeling The NRCs SE for NEI 06-09-A specifies that the 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. Table E1-1 of LAR Enclosure 1 identifies each LCO in the TSs proposed for inclusion in the RICT program. The table also describes whether the systems and components covered by the LCO are modeled in the PRA and, if so, presents both the design success criteria and PRA success criteria. For certain LCOs, the table explains that the associated SSCs are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some LCOs, the LAR did not provide an adequate description for the NRC staff to conclude that the PRA modeling will be sufficient.

a) Regarding TS LCO 3.3.5.1.B, Table E1-1, states that, for emergency core cooling system (ECCS) actuation instrumentation for core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection (HPCI), and diesel generators (DGs) that SSCs associated with Functions 3.a and 3.b are not explicitly modeled in the PRA. However, the associated Table E1-1 comments entry states that all of the SSCs can be explicitly modeled in the RTR tool. The NRC staff notes that the Functions 3.a and 3.b appear to be associated with HPCI instrumentation. It is unclear to the staff what system is associated with Functions 3.a and 3.b and if this associated instrumentation is incorporated into the RTR tool.

i.

Clarify the system(s) associated with TS LCO 3.3.5.1.B Functions 3.a and 3.b.

Include in this response if the actuation instrumentation related to these two functions are explicitly modeled in the RTR tool.

ii.

If the actuation instrumentation related to Functions 3.a and 3.b of TS LCO 3.3.5.1.B are not explicitly modeled in the RTR tool, then provide the following information:

a. Identify the RTR model surrogates to be used for Functions 3.a and 3.b of TS LCO 3.3.5.1.B

b. Provide justification that the surrogate(s) is related and bounds both Functions 3.a and 3.b of TS LCO 3.3.5.1.B.

b) Regarding TS LCO 3.3.5.1.C, Table E1-1, states that for ECCS actuation instrumentation for CS, LPCI), and DGs that SSCs associated with Functions 3.c and 3.g are not explicitly modeled in the PRA. However, the associated Table E1-1 comments entry states that all of the SSCs can be explicitly modeled in the RTR tool. It is unclear to the NRC staff what system is associated with Functions 3.c and 3.g and if the associated instrumentation is in the RTR tool.

i.

Clarify the system(s) associated with TS LCO 3.3.5.1.C Functions 3.c and 3.g.

Include in this response if the actuation instrumentation related to these two functions are explicitly modeled in the RTR tool.

ii.

If the actuation instrumentation related to Functions 3.c and 3.g of TS LCO 3.3.5.1.C are not explicitly modeled in the RTR tool, then provide the following information:

a. Identify the RTR model surrogates to be used for Functions 3.c and 3.g of TS LCO 3.3.5.1.C

b. Provide justification that the surrogate(s) is related and bounds both Functions 3.c and 3.g of TS LCO 3.3.5.1.C.

c) Regarding TS LCO 3.6.1.2.C, Table E1-1, states that for primary containment air locks not modeled, that a large pre-existing containment isolation failure that is modeled will be used as a surrogate. It is unclear to the NRC staff how pre-existing containment isolation failure is either conservative or bounding.

Provide justification that the surrogate conservatively bounds TS LCO 3.6.1.2.C.

d) Regarding TS LCO 3.6.1.3.A, Table E1-1, states that for primary containment isolation valves not modeled, that a large pre-existing containment isolation failure that is modeled will be used as a surrogate. It is unclear to the NRC staff how pre-existing containment isolation failure is either conservative or bounding.

Provide justification that the surrogate conservatively bounds TS LCO 3.6.1.3.A.

APLA Audit Question 08-Credit for FLEX Equipment and Actions Section 2 of Enclosure 9 of the RICT LAR states that FLEX (Diverse and Flexible Coping Strategies) is credited in the Quad Cities internal events PRA (FPIE model), which includes internal flooding and the FPRA.

NRC memorandum dated May 6, 20226, provides the NRCs staff updated assessment of identified challenges and strategies for incorporating FLEX equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.2007. The NRC staff states in conclusion 4 of the memo: Licensees that choose not to use the generic 6 U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments, dated May 6, 2022 (ML22014A084).

7 U.S. NRC, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, RG 1.200, Revision 3, December 2020 (ML20238B871).

failure probabilities in Pressurized-Water Reactor Owners Group (PWROG)-18042 to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRAs used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval.

It appears that NUREG-6928 fixed equipment failure rates, including a 2x increase, were used as probabilities for FLEX portable equipment. It is unclear to the NRC staff how the Quad Cities approach satisfies the concerns of Conclusion 4.

The enclosure explains how NRCs Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments (ML22014A084) is addressed in the modeling of FLEX.

Address the following:

a) Propose a mechanism to incorporate updated FLEX parameter values in accordance with PWROG-18042-N into the Quad Cities PRA models used for RICT calculations prior to implementing the RMTS program.

-OR-Alternatively, identify the LCO conditions impacted by the treatment of this modelling uncertainty for which RMAs will be applied during a RICT. Include discussion of the kinds of RMAs that would be applied and justification that the RMAs will be sufficient to address the modeling uncertainty.

b) Provide a discussion detailing the methodology used to assess operator actions related to installation and operation of FLEX equipment. The discussion should include:

i.

A list of the FLEX-related operator actions and a summary description of the plant-specific HRA used as the basis to develop the HEPs for each operator action. Include an evaluation of the HRA associated with declaration of Extended Loss of AC (alternating current} Power (ELAP).

ii. If the FLEX-related HRA is not in accordance with the NRC memorandum dated May 6, 2022, justification that the HRA assumptions have an inconsequential impact on the RICT calculations.

iii. If, in response to part ii) above, it cannot be determined that the cited assumptions have an inconsequential impact on the estimated RICTs, then identify the LCO conditions impacted by the treatment of this modelling uncertainty for which RMAs will be applied during a RICT. Include a discussion of the programmatic changes that the licensee will consider in order to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional RMAs and justification that they are sufficient to address the modeling uncertainty).

c) If the PRA modeling of FLEX equipment and/or operator actions is revised or updated to be in accordance with the NRC memorandum dated May 6, 2022, provide justification that the revisions do not meet the definition of an PRA upgrade as defined by RG 1.200.

-OR-Alternatively, if justification cannot be provided, propose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the PWROG FLEX equipment reliability modeling and/or EPRI FLEX HRA [Electric Power Research Institute/FLEX/human reliability analyses] methodology for the ANO -2 PRA models. Include in the mechanism to close out all Facts and Observations (F&Os) that result from the FSPR prior to implementing the RMTS program.

APLB Question 01 - Use of Unacceptable Methods The LAR provides the history of the FPRA peer review but does not appear to discuss methods used in the FPRA. Methods may have been used in the FPRA that deviate from guidance in NUREG/CR-6850, "EPRI/NRC-RES [Research]-Fire PRA [FPRA] Methodology for Nuclear Power Facilities," (ML052580075, ML052580118, and ML103090242), or other acceptable guidance (e.g., frequently asked questions (FAQs), NUREGs, or interim guidance documents).

a) Identify methods used in the FPRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance.

b) If such deviations exist, then justify their use in the FPRA and impact on the RICT program.

c) As an alternative to item b above, add an implementation item to replace those methods with a method acceptable to NRC prior to the implementation of the RICT program.

Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

APLB Question 02 - Reduced Transient Heat Release Rates (HRRs)

The key factors used to justify using transient fire reduced HRRs below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, NRC, to Biff Bradley, NEI, Recent Fire PRA Methods Review Panel Decisions and Electrical Power Research Institute (EPRI) 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires,"

(ADAMS Package ML12172A406).

If any reduced transient HRRs below the bounding 98 percent HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion:

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

b) 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.

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

APLB Question 03 - Treatment of Sensitive Electronics FAQ 13-0004, Clarifications on Treatment of Sensitive Electronics (ML13322A085) provides supplemental guidance for application of 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.

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 approach cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no impact on the RICT calculations.

c) As an alternative to item b above, add an implementation item to replace the current approach with an acceptable approach prior to the implementation of the RICT program.

Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

APLB Question 04 - Minimum Joint Human Error Probability (HEP/JHEP)

NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines-Final Report,"

(ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in HRAs. NUREG-1921 refers to Table 2-1 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA),"

(ML051160213), which recommends that joint HEP values should not be below 1E-5. Table 4-4 of EPRI 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, 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.

LAR Enclosure 9, Table E9-1, regarding joint human error probabilities (JHEPs) in the FPRA, states that the sensitivity study was performed since the Quad Cities implements a JHEP floor of 1E-06. The sensitivity study utilized the JHEP floor value of 1E-05, which is consistent with industry guidance. The LAR states that the sensitivity demonstrated that this source of uncertainty had a slight impact on FPRA results. However, during the staffs portal review of the Quad Cities FPRA Sensitivity Analysis Notebook, it was noted that the sensitivity results provided in Section 4.2.3 demonstrates a 9 percent impact in overall increase in CDF and LERF for Unit 1.

In addition, Section 8.8 of the Assessment of Key Assumptions and Sources of Uncertainty Notebook demonstrates a 14 percent impact on TS LCO 3.8.7.C.4 RICT calculation and provides results for only seven other TS LCO proposed for the RICT program. It is unclear to the NRC staff that this assumption has no impact on the RICT program. Therefore, address the following:

a) Provide justification, such as a RICT sensitivity study that the that the minimum joint HEP value has no impact on the remaining TS LCOs proposed for the RICT application.

b) If, in response part (a), if it cannot be justified that the minimum joint HEP value has no impact on the application, then provide the following:

i. Confirm that each joint HEP value used in the FPRA 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.

ii. If joint HEP values used in the FPRA below 1E-5 cannot be justified, add an implementation item to set these joint HEPs to 1E-5 in the FPRA prior to the implementation of the RICT program.

APLB Question 05 - Obstructed Plume Model NUREG-2178, Volume 1 Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume, (ML16110A14016) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction.

Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.

a) If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.

b) Justify any modelling in which the base of an obstructed plume is located at less than one half of the cabinet's height.

c) As an alternative to item b above, add an implementation item to remove credit for the obstructed plume model in the FPRA prior to the implementation of the RICT program.

APLB Question 06 - Well-Sealed Motor Control Center (MCC) Cabinets Guidance in FAQ 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of robustly or well-sealed. Thus, for cabinets of 440V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440V and higher, the original guidance in Chapter 6 remains and requires that Bin 15 panels which house circuit voltages of 440V or greater are counted because an arcing fault could compromise panel integrity (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 (ML15119A176) provides the technique for evaluating fire damage from motor control center (MCC) cabinets having a voltage greater than 440V. Therefore, propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440V or greater.

a) Describe how fire propagation outside of well-sealed MCC cabinets greater than 440V is evaluated.

b) If well-sealed cabinets less than 440V are included in the Bin 15 count of ignition sources, provide justification for using this approach as this is contrary to the guidance.

APLB Question 07 - Influence Factors for Transient Fires NUREG/CR-6850, Section 6, Fire Ignition Frequencies, and FAQ 12-0064 Hot Work/Transient Fire Frequency Influence Factors" (ML12346A488) describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

a) Indicate whether the methodology used to calculate hot work and transient fire frequencies applies influencing factors using NUREG/CR-6850 guidance or FAQ 12-0064 guidance.

b) Indicate whether administrative controls are used to reduce transient fire frequency, and if so, describe and justify these controls c) Indicate whether you have any combustible control violations and discuss your treatment of these violations for the assignment of transient fire frequency influence factors. For those cases where you have violations and have assigned an influence factor of 1 (Low) or less, indicate the value of the influence factors you have assigned and provide your justification.

d) If you have assigned an influencing factor of 0 to maintenance, occupancy, or storage, or hot work for any fire physical analysis units (PAUs) provide justification.

e) If a weighting factor of 50 was not used in any fire PAU, provide a sensitivity study that assigns weighting factors of 50 per the guidance in FAQ 12-0064.

APLB Question 08 - PRA Treatment of Dependencies between Units 1 and 2 Quad Cities, Units 1 and 2, are adjoined and thus have common areas. 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, common areas, and existence of shared systems. Therefore, address the following:

a) Explain how the risk contribution of fires originating in one unit is addressed for the other unit given impacts due to 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 and explain how the risk contributions of such scenarios are allocated in the LAR.

b) Explain how the contributions of fires in common areas are addressed including the risk contribution of fires that can impact components in both units.

c) Explain the extent to which systems are shared by both units and whether shared systems are credited in the PRA models for both units. If shared systems are credited in the PRA models for each unit, then explain how the PRAs address the possibility that a shared system is demanded in both units in response to a single initiating event or fire initiator.

APLB Question 09 - Fire Scenario Treatment of the Main Control Board (MCB)

Traditionally, the cabinets on front face of the MCB have been referred to as the MCB for purposes of FPRA. Appendix L of NUREG/CR-6850, (ML052580075) provides a refined approach for developing and evaluating those fire scenarios. FPRA FAQ 14-0008, Main Control Board Treatment, dated July 22, 2014 (ML14190B307) clarifies the definition of the MCB and effectively provides guidance for when to include the cabinets on the back side of the MCB as part of the MCB for FPRA. It is important to distinguish between MCB and non-MCB cabinets because misinterpretation of the configuration of these cabinets can lead to incomplete or incorrect fire scenario development. This FAQ also provides several alternatives to NUREG/CR-6850 for using Appendix L to treat partitions in an MCB enclosure. Therefore, address the following:

a) Briefly describe the main control room MCB configuration, and use the guidance in FAQ 14-0008, to determine whether cabinets on the rear side of the MCB are a part of the MCB. Provide your justification using the FAQ guidance.

b) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure using the definition in FAQ 14-0008, then confirm that guidance in FAQ 14-0008 was used to develop fire scenarios in the MCB and determine the frequency of those scenarios.

c) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure and the guidance in FAQ 14-0008 was not used to develop fire scenarios involving the MCB, then provide a description of how the fire scenarios for the backside cabinets are developed and an explanation of how the treatment aligns with NRC accepted guidance.

d) If in response to parts (c) above, the current treatment of the MCB and those cabinets on the rear side of the MCB cannot be justified using NRC accepted guidance, then justify that the treatment has no impact on the RICT calculations. Alternatively, propose a mechanism that ensures that the FPRA is updated to treat the MCB enclosure consistent with the guidance in FAQ 14-0008, prior to implementation of the RICT program.

APLB Question 10 - FPRA Methods for Outdated FPRA and Peer Review The LAR states, in part, the Internal FPRA model was developed consistent with NUREG/CR-6850 and only utilizes NRC approved methods. As part of the ongoing PRA maintenance and update process described in the LAR, the licensee will address Internal FPRA methods approved by the NRC since the development of the Internal FPRA. Furthermore, in the LAR, the licensee specifies that a full-scope FPRA model peer review was performed in 2013.

There have been numerous changes to the FPRA methodology since the last full scope peer review of the FPRA. The integration of NRC-accepted FPRA methods and studies described below that are relevant to this submittal could potentially impact the TSTF-505 results and/or the CDF and LERF. NRC has issued updated guidance for aspects of FPRA that supplant earlier guidance issued by NRC.

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

(ML16343A058) regarding the updated approach to credit incipient fire detections systems.

NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, (ML15016A069) regarding changes in fire ignition frequencies and non-suppression.

probabilities.

NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), Volume 2, Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure, (ML14141A129) regarding possible increases in spurious operation probabilities.

NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinets Fires in Nuclear Power Plants, (ML20157A148) regarding electrical cabinet fires.

NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), (ML20168A655) regarding heat release rates (Volume 2).

Section 2.5.5 of RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, provides guidance that indicates additional analysis is necessary to ensure that contributions from the above influences would not change the conclusions of the LAR.

a) Provide a detailed justification for why the integration of the above NRC accepted FPRA methods and studies would not significantly impact the RICT calculation. As part of this justification, identify potential FPRA methodologies used in the FPRA that are no longer accepted by the NRC staff. Provide technical justification for methods in Quad Cities FPRA not accepted by the NRC staff and evaluate the significance of their use on the RICT estimates.

-OR-b) Alternatively, if the above guidance has been implemented in Quad Cities FPRA, provide the following:

i.

Indicate whether the changes to the FPRA are PRA maintenance, or a PRA upgrade as defined in ASME/ANS [American Society of Mechanical Engineers/American Nuclear Society] RA-Sa-2009, Section 1-5.4, as qualified by RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML20238B871), along with justification for the determination.

ii.

Discuss the focused scope (or full scope) peer review(s) that was performed to evaluate the changes that were determined in Part b.i. above to constitute a PRA upgrade and provide the date for when the peer review(s) was performed and for when the peer review report(s) that evaluated the incorporation of the method(s) was approved.

APLB Question 11 - Open FPRA Facts and Observations (F&O)

RG 1.200, Revision 3, 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-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 F&Os recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the NEI guidance documents NEI 05-04, NEI 07-12, and NEI 12-13, titled NEI 05-04/07-12/12-06 Appendix X: Close-out of Facts and Observations (F&Os) (ML17086A431), which was accepted by the NRC in a letter dated May 3, 2017 (ML17079A427).

Section 4 of Enclosure 2 to the LAR states than one FPRA F&O, F&O 9-1, remains open and provides a succinct disposition. However, the LAR does not provide the full description of the finding from the FSPR, the recommendations from the FSPR team to address the finding, and a current disposition of this open F&O. Provide the FSPR full description, comments, recommendations, and licensee disposition related to this application for F&O 9-1.

APLC Audit Question 01 - Darley Pump Section 2.3.1, Item 7 of NEI 06-09, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06-09, states that Where PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

The revised flood analysis refers to a portable Darley pump to be utilized to provide makeup flow during certain external flooding scenarios. QCOA 0010-16, Flood Emergency Procedure, Revision 29, states that to provide a suction source for the Darley pump while waters are between 595 feet (ft) and 599 ft elevation, a hose should be routed over the 4 ft barrier installed in the Reactor Building (RB) 1/2 trackway personal access. Access to flood water suction provides Operations with an additional method of cooling makeup before flood waters exceed 599 ft.

However, this sequence does not appear to be discussed in the estimate of flood risk in the TSTF-505 LAR. The LAR provides an estimate of 0.3 CCDP (conditional core damage probability) for operators to install flood barriers.

a) Describe how the Darley pump mitigates risk during the external flood scenarios.

b) Provide the risk contribution (importance) of this sequence.

c) Describe the training operators receive regarding this scenario, including the frequency of training.

d) Describe how the timing for taking the action will be validated during training and how deviations between the actual and modeled required time will be addressed.

e) Describe credit given for this scenario.

f) Clarify if the licensee intends for this to be an implementation item for the applications.

Provide, if needed, a description of the implementation item.

APLC Audit Question 02 - External Flooding Procedures The TSTF-LAR, Attachment 4, Section 5, states that completion of EC 636914 Update to LIP Barriers to Assist the Station External Flood Response, which is being developed to modify barriers to protect the plant up to the 599 ft elevation, is tracked as a RICT Program Implementation Item. It appears that EC 636912, Update to the Station External Flood Response to Support Risk Reduction, as well as changes to QCOA 0010-16 are not complete. The LAR states that the addition of the LIP barrier installation to QC 0010-16 with available time will greatly reduce the risk to the station from external river floods.

Clarify why the completion of these documents are not proposed as implementation items.

Alternatively, provide implementation items that ensure completion prior to implementing either application.

APLC Audit Question 03 - Equipment Operators and External Flooding Barriers The LAR states that installation of the local intense precipitation (LIP) barriers in the to-be-revised QCOA-0010-16 requires 8 equipment operators working concurrently. QCOA 0010-22 states that Fastlogs weigh 113 lbs (pounds) and installation of each Fastlog requires a minimum of two people. The new flooding analysis states that operators will need additional training to support the flood strategy response where barriers are installed to prevent excessive water intrusion for floods up to 599 ft.

a) Clarify if licensee procedures account for the number of required equipment operators on site to perform this installation.

b) Provide the frequency of training that the operators receive or will receive on the to-be-revised procedure regarding this installation.

c) Describe how the timing for taking the action will be validated during training and how deviations between the actual and modeled required time will be addressed.

d) It appears that the HEP/CCDP for this installation incorporates the failure of the barriers themselves.

i.

Describe the quality control for these barriers.

ii. Provide the barriers storage location.

iii. Describe and justify the barriers failure rates.

APLC 03 Addendum The licensee presentation on external flooding on January 23, 2024, listed seven HFEs, six with a HEP value of 5E-02 and one with a 3E-01 value. The 5E-02 value is stated as a conservative value assigned to HEP-1A (Tsw = 45 minutes), HEP-1B (Tsw = 45 minutes), HEP-3 (Tsw = 5 minutes), HEP-8 (Tsw = 25 minutes), HEP-9 (Tsw = 40 minutes), and HEP-11 (Tsw = 45 minutes). [Tsw represents the amount of time from the initiating event (T = 0) to complete the action.] It is unclear to the NRC staff how HFEs with significantly different Tsw values and represent three types of installations (swing gates, panels, and fastlogs) would have the same HEP value. It is unclear what HFE XF-LIP-HEP represent that has a Tsw value of 45 minutes and a HEP of 3E-01.

Section 5.3.3 of NUREG 1792 provides good HRA practices for post-initiator HFEs. Good Practice #2 states no HFE screening value should be lower than 0.1 or lower than the worst-case anticipated value (which appears to be 3E-01). Good Practice #4 states to revisit the use of post-initiator screening values versus detailed assessments for special applications. It is unclear to the NRC staff if the above HFEs are developed or screened and if the assigned HEP value is appropriate and if screened that these HFEs should be adequately assessed. The NRC staff notes that Capability Category I of the 2009 ASME/ANS PRA Standard SR HR-G1 the use of screening values for HECC-I states to use conservative estimate for the HEPs of the HFEs. The presentation represented these values as conservative. It is unclear to the staff if the HRA development of these HFEs are conservative or bounding.

a) Clarify what operator action HFE XF-LIP-HEP represents and why its probability is different than the other six HFEs.

b) Clarify if the seven HFEs listed associated with the barriers are screened or developed operator actions. For those HFEs that are screened, justify that they are nonsignificant operator actions.

c) For the HFEs that have a screening value, that constitute different actions, and have different Tsw values, justify that the same screening value is appropriate.

d) Provide justification that the HEP value of 5E-02 is consistent with the guidance of NUREG-1792.

e) Provide justification that these HEP values are conservative.

Good Practice #6 of NUREG-1792 for post-initiator state to account for dependencies among post-initiator HFEs. SR HR-G7 requires, for multiple actions in the same accident sequence, to assess the degree of dependence and to calculate a joint HEP. It is unclear to the staff if a dependency analysis (DA) was performed for the external flood sequence. The presentation appears to state that the following actions would also be performed during this sequence:

Remove decay heat Install residual heat removal (RHR) (6 fire hose cross-ties to fire water supply Fill both torus Remove shiel plugs, drywell heads, and reactor vessel head Set up portable pump (Darley) for decay heat removal Fill Radwaste tanks with fire system water Portable makeup demineralizers to the CST (condensate storage tank)

Fill reactor cavities and separator-dryer pools Remove gates between storage pools Rack out all main breakers for equipment below 608 feet (lose normal decay heat removal systems)

Open plant doors The NRC staff understands that the licensees analysis did not credit the portable pump, however, it is unclear if the other actions listed above would be performed in addition to the barrier installations. The NRC staff notes that other tasks may need to be performed related to plant operations during the barrier installation. It is unclear to the staff if all of the relevant operator actions were included in the analysis. If the licensees analysis relies on the arrival of offsite personnel, the NRC staff notes that access to the plant (flooded roads and bridges) should be taken into account.

f) Clarify if a DA was included in the external combined effects flood.

g) Clarify what other operator actions would be required during the installation of the barriers. Include in this discussion their inclusion in the DA.

h) Clarify if offsite personnel are required for this response. Include in this discussion if plant access was taken into account.

i)

If a DA was excluded from the analysis, provide justification for its exclusion.

j)

Based on the responses to parts (d), (e), and (i) provide justification that sum of these issues does not significantly impact the application.

APLC Audit Question 04 - RB Crane and External Flooding EC 636914 states that the RB crane will be loaded onto the emergency diesel generator (EDG) if a LOOP occurs during external flooding.

a) Describe the function of the RB crane used in this scenario.

b) Is loading the RB crane onto the EDG proceduralized as part of the EDGs loading following a LOOP? If not, how will the licensee ensure that such loading is performed in practice.

APLC Audit Question 05 - Revised External Flooding Evaluation and FLEX Integrated Assessment The NRC staff reviewed the licensees integrated assessment (IA;) for the revised flood hazard as part of the agencys post-Fukushima actions. The staffs review is documented at ML19168A196.

a) Identify instances where the revised flooding analysis and actions provided for this application change the information provided to the staff as part of the licensees IA.

b) If changes to the licensees IA are identified, for each change, justify why the staffs conclusions on the IA continue to remain valid.

c) Clarify whether the revised flooding analysis and actions provided for this application change the licensees demonstration of compliance with 10 CFR 50.155.

APLC Audit Question 06 - 10 CFR 50.69 LAR Hazard Screening Process NEI 00-048, Figure 5-6, provides guidance to be used to determine SSC safety significance. The same document states, in part, that if it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the Licensing Support System (LSS) category.

In Section 3.2.4 of the Quad Cities 10 CFR 50.69 LAR, the licensee states that All external hazards, except for seismic, were screened for applicability to Quad Cities using a plant-specific evaluation in accordance with Generic Letter 88-20 (Reference [27]) and updated to use the criteria in ASME PRA Standard RA-Sa-2009.

Likewise, Attachment 4 of the Quad Cities 10 CFR 50.69 LAR lists all external hazards as screened except for seismic hazard with an alternate approach. The guidance in NEI 00-04, Figure 5-6, regarding SSCs that play a role in screening a hazard is not discussed in Section 3.2.4 nor in Attachment 4 of the LAR. Therefore, it appears to the NRC staff based on this lack of information that at the time an SSC is categorized it will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard because that evaluation has already been made.

NRC staff notes that plant changes, plant or industry operational experience, updates to hazard frequency information, and identified errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard.

a)

Clarify whether or not an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard.

b)

If an SSC will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard at the time of categorization because that evaluation has already been made, then explain how plant changes, plant or industry operational experience, updated information in hazard frequencies, and identified errors or limitations that could change that decision are addressed.

In the table in Attachment 4 of the Quad Cities LAR under the external flooding evaluation, the license states Table A4-1 includes a list of LIP barriers credited for screening this hazard. Table A4-1 lists 14 barriers credited for screening the external flood hazard (six of which require manual action). In the Quad Cities TSTF-505 LAR, the licensee states that the addition of the LIP barrier installation to QC 0010-16 with available time will greatly reduce the risk to the station from external river floods.

c)

Describe how the SSCs listed in Table A4-1 are to be categorized using the guidance in NEI 00-04, Figure 5-6.

8 NEI 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline", July 2005 (ML052910035).

APLC Audit Question 07 - Chemical Complex Evaluation NEI 00-04, Revision 0, Section 5, Component Safety Significance Assessment states, If the plant does not have an external hazards PRA, then it is likely to have an external hazards screening evaluation that was performed to support the requirements of the IPEEE [individual plant examination of external events]. NEI 00-04, Revision 0, also states in Section 3.3.2, Other Risk Information (including other PRAs and screening methods), that the characterization of the adequacy of risk information should include a basis for why the other risk information adequately reflects the as-built, as-operated plant.

The TSTF-505 and 10 CFR 50.69 LARs, Table E4-16, and Attachment 4, respectively, state under the Industrial or Military Facility section that none of the operations at Cordova Industrial Park pose any threat to Quad Cities from explosion, explosive shock, resulting missiles, or toxic fumes release yet there is no mention of the effect of the CF Industries Chemical complex on the plant. Please explain the impact of the CF Industries Chemical complex on the plant and either justify that the impact can be screened for this application or describe how the impact is included in the RICT program.

APLC Audit Question 08 - Screening of Snowfall Risk NEI 00-04, Revision 0, Section 5, Component Safety Significance Assessment states, If the plant does not have an external hazards PRA, then it is likely to have an external hazards screening evaluation that was performed to support the requirements of the IPEEE. NEI 00-04, Revision 0, and also states in Section 3.3.2, Other Risk Information (including other PRAs and screening methods), that the characterization of the adequacy of risk information should include a basis for why the other risk information adequately reflects the as-built, as-operated plant.

The TSTF-505 and 10 CFR 50.69 LARs, Table E4-16, and Attachment 4, respectively, states that criterion C4 (event is included in the definition of another event) and criterion C5 (event develops slowly, allowing adequate time to eliminate or mitigate the threat) was used to screen the snow hazard. The LAR focuses on potential flooding impacts, but not on the design basis roof live load or the maximum recorded snowfall for the site. It is unclear to the NRC staff whether the risk of this hazard is adequately considered for this application.

Justify the screening of risk associated with snowfall from the application (e.g., by comparing historical maximum snowfall against the design basis, showing that the occurrence frequency of snowfall events that could challenge the plant is low).

Probabilistic Food Hazard Analysis [PFHA] Audit Questions 1 through 5 are related to Appendix A to the PFHA Report. Quad Cities Multi-Site Stochastic Daily Weather Simulation Report PFHA Audit Question 01 - Section 2.0, Study Location a) Provide discussions on the potential reduction in spatial variability of daily precipitation and temperature fields resulting from the selection of 49 gauging stations out of 250 and further simplification to 10 subbasins covering the basin area of approximately 88,600 square miles. Additionally, elaborate discussions on the potential uncertainties introduced by utilizing the simplified weather fields represented by 10 subbasins for hydrologic and flood frequency analyses.

b) Explain the methodology employed to handle missing weather data, if any, in the stochastic modeling.

PFHA Audit Question 02 - Section 3.0, Regional Stochastic Weather Generation a) This study uses the daily RMAWGEN model with monthly mean weather variables (i.e.,

station mean, climate mean and/or climate projection means) to generate stochastic daily weather scenarios. Expand the descriptions on the approach utilized to maintain the daily temporal variability in historical climate data while using monthly means in the RMAWGEN model.

b) (Figure 2.c): The plot shows that, unlike the minimum and maximum temperatures, the precipitation data are not distributed normally even after the Gaussian transformation.

Discuss the adequacy of the RMAWGEN modeling in addressing transformed non-normal precipitation data and its potential impact on low flow frequency analysis.

c) (Section 3.1): The licensee generated synthetic weather data using a general linear model in RMAWGEN following Wilks' approach for spatial correlation. Describe the process and assumptions of the Wilks approach (or RGENERATEPREC) applied to the licensees stochastic modeling, supplemented by the result of calibration and data generation to demonstrate that the generated data preserve the statistics (i.e., wet-dry spells and size and duration of storms) of measured weather data.

PFHA Audit Question 03 - Section 3.2, Precipitation and Temperature Model Calibration a) Provide an explanation of calibration of the multi-variate auto-regression (VAR) models. If available, this explanation could include selection of the order (time-lag) of VAR models, utilization of exogenous variables, goodness-of-fit statistics, and variance of white noises.

b) Section 4.0 says the VAR models were calibrated for normal-, wet-, and dry-year data, while Sections 3.2 and 6.1 state that the VAR models were calibrated with observed 1949-2019 data. Clarify the discrepancy. In cases where the VAR models were calibrated for different years, describe how the calibrated parameters are different for different year calibration sets, and how they were merged to the final VAR models for generating stochastic weather data.

c) (Figure 3, Q-Q Plots): The plots show notable dispersions on high precipitation values at most subbasins. Discuss the implications of data dispersions in generated daily precipitation depths on flood frequency analysis at the 1E-6 annual exceedance probability (AEP) level.

PFHA Audit Question 04 - Long-term Calibration a) Unlike its title, this subsection describes spatial distribution and subbasin magnitude of generated precipitations for three selected reference years. If available, provide detailed descriptions on the calibration and validation of SPAR and PRISM dataset.

b) The VAR model for precipitation relies only on gauging data observed within the watershed basin. The precipitation estimates could be more conservative if the licensee adopts data observed both inside and outside (homogeneous area) of the basin using the concept of the stochastic storm transposition. Discuss the potential impacts of adding historical data recorded outside the basin on evaluating low-frequency floods and their uncertainties.

PFHA Audit Question 05 - Section 6, Climatic Change a) This study analyzes climatic change projection scenarios based on the Intergovernmental Panel on Climate Change (IPCC), fifth assessment report (Fifth Assessment Report (AR5): Representative Concentration Pathways (RCPs) 4.5 and 8.5). Discuss the potential impacts of utilizing site-specific information from the latest climate report (AR6) for the 2015~2023 period or corresponding climate projection maps (e.g., Coupled Model Intercomparison Project Phase 6 (CHIP6), if available) on the flood frequency estimates in the region.

b) This study tested five climate scenarios: normal condition, moderate and severe scenarios with and without evapotranspiration input. The IPCC AR6 summarizes that terrestrial global average precipitation has increased, as has the frequency of heavy precipitation events. Explain the rationale behind excluding certain climate scenarios (e.g., precipitation projections) from flood frequency analysis, despite evidence of increased heavy precipitation events in the IPCC AR6 report.

c) This study performs climatic change projections using the regional downscaled model output driven by the emission scenarios modeled by CMIP5 global climate model. Provide details on the climatic change projections using regional downscaled model output, including the temporal resolution of projection data and downscaled daily implementation.

d) It is not clear how the output of RCPs 4.5 and 8.5 are used to generate 12 future 1,000-year climate scenarios, and how the climate scenarios (monthly?) are used in the VAR daily temperature model. Clarify the stochastic modeling and calibration processes for generating future weather scenarios, particularly addressing the incorporation of RCPs 4.5 and 8.5 scenarios and any associated uncertainties. Describe the process of treating different time intervals from monthly to daily in the context of climatic change projections in the calibrated VAR models.

e) Discuss how notable precipitation depth increases (esp., during the spring season) for climate change scenarios, as depicted in Figure 10, were incorporated into the precipitation VAR model.

PFHA Audit Questions 6 through 9 are related to the Flood Frequency Analysis for the Mississippi River PFHA Audit Question 06 - Section 4.3, Hydrologic Model Calibration a) For the continuous simulation, the licensee recalibrated the flood hazard reevaluation report HEC-HMS model with additional elements of canopy, linear reservoirs, etc. Briefly describe the refinement (change) of key model parameters obtained from the recalibration, emphasizing their impacts on estimating peak and time to peak flow events.

b) Discuss the adequacy of utilizing the 1991 calibration parameters for data generation and subsequent flood frequency analyses at the 1E-6 AEP level, considering the underestimation of peak and time to peak estimates as appeared in Figure 9 and biases (underestimations) in streamflow estimates for verification years (1988, 1993) as presented on Table 5.

PFHA Audit Question 07 - Section 5, Flow Frequency Analysis The data in Table 7 does not match the plotted points in Figure 15. Clarify the discrepancy.

PFHA Audit Question 08 For Figure 13 and Table 6, provide the mean flood frequency curve with its uncertainty bounds (i.e., 5th and 95th percentile bounds) to better understand the uncertainties associated with weather data generation and hydrologic modeling. This issue should be addressed to the curves provided in Section 7 (Uncertainty Analysis). When considering the flood frequency curve presented in Figure E4-3 of the TSTF-505 LAR, examine how the notable uncertainties in estimating flood frequency impact the suggested risk-informed flood protection or mitigation measures.

PFHA Audit Question 09 - Software Tools Ensure that the electronic model input/output files for HEC-HMS and stochastic weather generation models for calibration and key scenario simulations are available to the NRC staff for auditing purpose, as needed. Also, conform that the special software tools used in this study (e.g., the script programs developed for fitting frequency curves with large-size data) meet the requirements outlined in 10 CFR, part 50, Appendix B, Quality Assurance Criteria.

ML24110A049 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DRA/APLA/BC NRR/DRA/APLB/BC NAME RKuntz SRohrer RPascarelli SMakor (A)

DATE 4/18/2024 4/24/2024 5/3/2024 5/6/2024 OFFICE NRR/DRA/APLC/BC NRR/DEX/EEEB/BC NRR/DEX/EICB/BC NRR/DEX/EXHB/BC NAME SVasavada WMorton FSacko BHayes DATE 5/2/2024 5/3/2024 5/2/2024 5/2/2024 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME SMehta JWhited SWall for RKuntz DATE 5/3/2024 5/24/2024 5/28/2024