ML20167A184

Request for Additional Information Related to Seabrook License Amendment Request Regarding TSTF-411 and TSTF-418 (L-2019-LLA-0237)

ML20167A184
Person / Time
Site:	Seabrook
Issue date:	06/11/2020
From:	Justin Poole Plant Licensing Branch 1
To:	Browne K, Thomas C NextEra Energy Seabrook
Poole J, NRR/DORL/LPLI, 415-2048
References
EPID L-2019-LLA-0237
Download: ML20167A184 (5)

Text

From: Poole, Justin To: Browne, Kenneth; Thomas, Christine Cc: Danna, James

Subject:

Request for Additional Information Related to Seabrook License Amendment Request Regarding TSTF-411 and TSTF-418 (L-2019-LLA-0237)

Date: Thursday, June 11, 2020 4:20:00 PM Attachments: L-2019-LLA-0237 Final RAIs .pdf Ken/Christine, By letter dated November 1, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19310D804), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted a license amendment request (LAR) to adopt Technical Specification Task Force (TSTF) travelers: TSTF-411, Surveillance Test Interval Extension for Components of the Reactor Protection System (WCAP-15376), and TSTF-418, RPS and ESFAS Test Times and Completion Times (WCAP-14333), for Seabrook Station, Unit No. 1. In reviewing the submitted information, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is necessary to complete its review.

On May 15, 2020, the NRC staff sent NextEra the DRAFT RAIs to ensure that the questions are understandable, the regulatory basis is clear, there is no proprietary information contained in the RAI, and to determine if the information was previously docketed. On June 11, 2020, the NRC and NextEra held a clarifying call. During the call, NextEra requested a response date of 30 days from the date of this email. The NRC staff informed Exelon that this timeframe is acceptable. The attached is the final version of the RAIs. These RAIs will be put in ADAMS as a publicly available document.

Justin C. Poole Project Manager NRR/DORL/LPL I U.S. Nuclear Regulatory Commission (301)415-2048

REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUESTS TO ADOPT TSTF-411 AND TSTF-418 NEXTERA ENERGY SEABROOK LLC, SEABROOK STATION, UNIT NO. 1 RAI 01 - PRA Maintenance and Upgrades Table 3.2.1-1-f of the license amendment request (LAR) provides a list that summarizes the significant probabilistic risk assessment (PRA) model changes and core damage frequency (CDF) impacts for the Seabrook internal events PRA (IEPRA) that includes internal floods. The U.S. Nuclear Regulatory Commission (NRC) staff observes that there have been several model updates since the last peer reviews were performed (e.g., complete revision of latent human failure event analysis and complete revision of the internal floods analysis in 2011). It is not clear whether these PRA changes are considered PRA upgrades or PRA maintenance in accordance with the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as endorsed by Regulatory Guide (RG) 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML090410014). Furthermore, it is not clear for the updates incorporated into the IEPRA (includes internal floods) whether these upgrades, if any, have been peer reviewed or align to the scope of a specific peer review(s) provided in Section 3.4 of the LAR.

Based on the observations provided above, address the following:

a. For the PRA model changes that occurred in 2011 and 2014 provided in LAR Table 3.2.1-1-f, provide sufficient justification to ascertain whether each of the model changes constitutes a PRA maintenance or upgrade (i.e., describe the change and include justification) as defined in the ASME/ANS RA-Sa-2009 PRA standard, as qualified by RG 1.200, Revision 2. For each change, indicate whether the change was determined to be PRA maintenance or a PRA upgrade consistent with the endorsed PRA standard.

b. For any PRA upgrades identified in Part (a) above, either:

i. Identify any focused-scope (or full-scope) peer reviews that were performed in accordance with RG 1.200, Revision 2, to address each PRA upgrade. Provide the findings of these peer reviews and the associated dispositions as it pertains to the impact on this LAR if not already provided in the submittal dated November 1, 2019 (ADAMS Accession No. ML19310D804).

ii. Alternatively, if a peer review was not performed to address the upgrade(s), provide sufficient information for the NRC staff to compare the technical adequacy of the upgrade to RG 1.200, Revision 2, or provide a bounding or sensitivity evaluation of its effect to demonstrate that it has no impact on the conclusions of this risk-informed application.

RAI 02 - Open Peer Review Findings and Self-Assessment Items Tables 3.4.1 and 3.4.2 of the LAR provide the dispositions and resolutions for the Facts and Observations (F&Os) and self-assessment items that remain open following the peer reviews and Independent Assessment performed for the IEPRA. The disposition/resolution for F&O HR-E3-1 does not provide sufficient justification for the staff to discern that there is no impact for the requested technical specification (TS) change or that the licensee has resolved the F&O to meet capability category (CC) II for the associated supporting requirement (SR) of ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2.

For the disposition of F&O HR-E3-1, the licensee stated that a comprehensive review of all post-initiator dynamic operator actions associated with scenarios initiated at-power was performed by a former Seabrook Station operations shift manager and that only a documentation change is needed to close this finding and not expected to have any impact on the human error probabilities (HEPs) or PRA results for this risk-informed application.

Furthermore, the independent review team concluded that the action taken (i.e., comprehensive review performed by the former personnel) to address this finding does not fully meet the intent of the SR.

Procedures and operational practices continue to be revised and modified, therefore, the NRC staff interprets the intent of SR HR-E3 is to be performed by a current plant personnel to confirm that the interpretation of the procedures remains valid and is consistent with any revised or new plant observations and training procedures.

EITHER:

i. Provide justification (e.g., results and description of a sensitivity study performed) to confirm that this F&O has no impact on the conclusions of this risk-informed application.

OR ii. Perform the review and talk throughs with current plant operations and training personnel and update the human reliability analysis and IEPRA as necessary to meet the intent of SR HR-E3 at CC II.

RAI 03 - Reactor Coolant Pump Seal Modeling Section 2.5.1.2 of RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (ADAMS Accession No. ML17317A256) states, [t]he use of models for specific events or phenomena supports the development of the PRA model. In many cases, the industrys state of knowledge is incomplete, and opinions may vary on how the models should be formulated. Examples include approaches to modeling human performance, CCFs [common cause failures], and reactor coolant pump seal behavior upon a loss of seal cooling. This gives rise to model uncertainty. Section 2.5.1.2 of RG 1.174, Revision 2, states, [t]he impact of using alternative assumptions or models may be addressed by performing appropriate sensitivity studies or by using qualitative arguments, based on an understanding of the contributors to the results and how they are impacted by the change in assumptions or models. In addition, Section 2.5.2 states, [i]n general, the results of the sensitivity studies should confirm that the guidelines are

still met even under the alternative assumptions (i.e., change generally remains in the appropriate region).

In Table 3.2.1-1-f of the LAR, NextEra provides a summary of significant model changes and CDF impacts that include an update of the Reactor Coolant Pump (RCP) seal loss of coolant accident (LOCA) model that is credited in the IEPRA. Provide the following information to validate and confirm the PRA acceptability of the RCP seal LOCA model used in the risk evaluation performed to support the requested risk-informed application.

a. Provide a summary of the update to the RCP seal LOCA model, addressing the following aspects:

i. Describe any RCP seal modifications at Seabrook Station (e.g., installation of Westinghouse Generation III shutdown seals) associated with the updated RCP seal LOCA model.

ii. Describe the PRA update of the RCP seal LOCA model.

iii. Discuss the credit taken for any RCP seal modifications and the technical basis for that credit (e.g., technical report approved by the NRC). Demonstrate how any limitations and conditions delineated in the applicable NRC-approved guidance are being met [e.g., if the Westinghouse Generation III shutdown seals were installed and credited, those limitations and conditions in Section 3 of Topical Report PWROG-14001-P, Revision 1, "PRA Model for the Generation III Westinghouse Shutdown Seal," and Section 5 of NRC safety evaluation (SE) for PWROG-14001-P (ADAMS Accession No. ML17200A116)].

iv. Indicate, and provide justification, whether the updated RCP seal LOCA model is PRA maintenance or PRA upgrade, as defined in Section 1-5.4 of ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2. This discussion should be of sufficient detail to allow NRC staff to independently assess whether this change is a PRA maintenance or PRA upgrade (e.g., summarize the original method in the PRA and the new method, summarize the impact that this change has on significant accident sequences or the significant accident progression sequences).

b. If the PRA update of the RCP seal LOCA model is considered a PRA upgrade and a peer review(s) was performed for this upgrade, then discuss this peer review(s). In this discussion, describe the peer review process applied to the model; identify the guidance used to perform this peer review(s) (e.g., ASME/ANS RA-Sa-2009, NEI 05-04, RG 1.200, Revision 2); include any necessary gap- or self-assessments if current endorsed guidance/standards were not used in the peer review(s); provide all F&Os characterized as findings from the peer review(s) and the associated dispositions as it pertains to this application.

c. If the PRA update of the RCP seal LOCA model is considered a PRA upgrade and a peer review was not performed for this upgrade, then perform an appropriate sensitivity and/or bounding analysis (e.g., remove credit for RCP shutdown seals) that assesses the contribution of risk for the risk-informed application. Discuss this sensitivity/bounding analysis and provide updated risk values from the LAR to assess the risk impact.

Confirm that the results of this analysis still meet the acceptance guidelines in RG 1.174,

Revision 2, WCAP-14333-P-A, WCAP-15376-P-A, and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (ADAMS Accession No. ML100910008). If the acceptance guidelines are exceeded, then provide qualitative or quantitative arguments, based on an understanding of the contributors to the results and how they are impacted by the change in assumptions or models, to support the conclusions of the LAR. This discussion should include which metrics are exceeded and the conservatisms in the analysis and the risk significance of these conservatisms.

RAI 04 - Treatment of Key Assumptions and Key Sources of Uncertainty Section 3.3.2 of RG 1.200, Revision 2, provide guidance on how to identify, characterize and treat key sources of uncertainty relevant to a risk-informed application. Additionally, Section 3.3.2 of RG 1.200, Revision 2, defines key assumptions and key sources of uncertainty to be considered for relevance to the risk-informed application.

The licensee further states in Section 3.4 of the LAR, there are no sources of uncertainty that have an impact on the internal events and internal flooding quantitative risk model and thus there is no impact on the proposed application The NRC staff requests the following information to confirm the key assumptions and key sources of uncertainty were properly assessed from the base IEPRA (includes internal floods) model that has received peer reviews:

a. Provide a description of the process used to determine the key sources of uncertainty and key assumptions for the IEPRA (includes internal floods) model used to support this application.

i. A description of how key assumptions and key sources of uncertainty were assessed from the initial comprehensive list of PRA model(s) (i.e., base model) sources of uncertainty and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns determined to be applicable for this risk-informed application.

ii. A discussion on how the process and the criteria used to identify an assumption or source of uncertainty as key is consistent with RG 1.200, Revision 2, or other NRC-accepted methods.

b. If the process of identifying the key assumptions or key sources of uncertainty for the IEPRA model used to support this application cannot be justified for use in this risk-informed application, provide the results of an updated assessment that includes a description of each key source of uncertainty or key assumption identified along with the disposition (e.g., sensitivity study, etc.) to address the impact of each on this risk-informed application. Also, provide a description of the process used to determine the key sources of uncertainty and key assumptions in the updated assessment and how this process is consistent with RG 1.200, Revision 2, or other NRC-accepted methods.