ML20136A010

NRR E-mail Capture - Final RAIs Regarding Surry Units 1 & 2 LAR to Adopt 10 CFR 50.69

ML20136A010
Person / Time
Site:	Surry
Issue date:	05/14/2020
From:	Vaughn Thomas NRC/NRR/DORL/LPL4
To:	Geoffrey Miller Dominion Generation
References
Download: ML20136A010 (13)

Text

From:

Thomas, Vaughn Sent:

Thursday, May 14, 2020 3:07 PM To:

gary.d.miller@dominionenergy.com

Subject:

Final RAIs Regarding Surry Units 1 & 2 LAR to Adopt 10 CFR 50.69 Attachments:

Final RAIs Related to Surry 50.69 LAR_May 14 2020.docx

Gary, The attached file constitutes the staffs issuance of the final RAIs to Surry pertaining to the subject LAR. Per our clarification call today, May 14, 2020, the licensee agreed to provide its responses to the staff RAIs within 45 days from todays date, May 14, 2020 (due June 29, 2020).

Thank you, Vaughn Thomas, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Vaughn.Thomas@nrc.gov 301-415-5897

Hearing Identifier:

NRR_DRMA Email Number:

587 Mail Envelope Properties (MN2PR09MB5980EC0D89CED70309BD5D6EF4BC0)

Subject:

Final RAIs Regarding Surry Units 1 & 2 LAR to Adopt 10 CFR 50.69 Sent Date:

5/14/2020 3:06:43 PM Received Date:

5/14/2020 3:06:00 PM From:

Thomas, Vaughn Created By:

Vaughn.Thomas@nrc.gov Recipients:

"gary.d.miller@dominionenergy.com" <gary.d.miller@dominionenergy.com>

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MN2PR09MB5980.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 565 5/14/2020 3:06:00 PM Final RAIs Related to Surry 50.69 LAR_May 14 2020.docx 91889 Options Priority:

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REQUESTS FOR ADDITIONAL INFORMATION REGARDING THE SURRY POWER STATION, UNITS 1 AND 2, LICENSE AMENDMENT REQUEST FOR APPLICATION TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURE, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS" DOCKET NOs. 50-280 and 50-281 (EPID: L-2019-LLA-0269)

By letter dated December 6, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19343A019), Virginia Electric and Power Company (Dominion Energy Virginia, the licensee) submitted a license amendment request (LAR) for the Surry Power Station Units 1 and 2 (SPS). The proposed license amendment would modify the SPS licensing basis, by the addition of a license condition, to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The U.S. Nuclear Regulatory Commission (NRC) staff from Probabilistic Risk Assessment Licensing Branches A and C (APLA, APLC) have reviewed the LAR and request additional information (RAI) in order to complete the review.

RAI 01 - Proposed License Condition (APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant.

The guidance in Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ADAMS Accession No. ML052910035), allows licensees to implement different approaches, depending on the scope of their probabilistic risk assessment (PRA) (e.g., the approach if a seismic margins analyses is relied upon is different and more limiting than the approach if a seismic PRA is used). Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, (ADAMS Accession No. ML061090627), states, [a]s part of the NRC's review and approval of a licensee's or applicant's application requesting to implement 10 CFR 50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the licensee's categorization approach.

Section 2.3 of the LAR proposed the following license condition:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; the Appendix R program to evaluate fire risk; qualitative assessments of seismic insights; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 [American Society of Mechanical Engineers/American Nuclear Society (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 specified in License Amendment No. [XXX] dated [AMENDMENT DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from an Appendix R program fire risk evaluation to a fire probabilistic risk assessment approach).

The proposed license condition does not appear to reflect the alternate seismic approach provided in the LAR, which is a modified version of the Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, Tier 1 approach. In light of this observation, the licensee is requested to provide a revised license condition that refers to the alternate seismic approach provided in the application and any subsequent supplements.

RAI 02 - Overlap of Functions and Components (APLA)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant. The guidance in NEI 00-04 allows licensees to implement different approaches, depending on the scope of their PRA.

Section 7.1 of NEI 00-04 states, "[d]ue to the overlap of functions and components, a significant number of components support multiple functions. In this case, the SSC, or part thereof, should be assigned the highest risk significance for any function that the SSC or part thereof supports." Section 4 of NEI 00-04 states that a candidate low-safety-significant (LSS) SSC that supports an interfacing system should remain uncategorized until all interfacing systems are categorized. The LAR does not discuss consideration or implementation of the guidance in Section 7.1 of NEI 00-04.

The licensee is requested to explain how the categorization process will be implemented to ensure that the cited guidance in NEI 00-04 will be followed and that any functions/SSCs that serve as an interface between two or more systems will not be categorized until the categorization for all of the systems that they support is completed and that SSCs that support multiple functions will be assigned the highest risk significance for any of the functions they support.

RAI 03 - Crediting of FLEX in the PRA Model (APLA)

Paragraphs (c)(1)(i) and (ii) of 10 CFR 50.69 require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience. The NRC memorandum dated May 30, 2017, "Assessment of the Nuclear Energy Institute 16-06, 'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis" (ADAMS Accession No. ML17031A269), provides the NRC staffs position concerning incorporating mitigating strategies (FLEX) into a PRA in support of risk-informed decision making in accordance with the guidance in RG 1.200, Revision 2 (ADAMS Accession No. ML090410014).

To complete the staffs review of the FLEX strategies modeled in the PRA, the licensee is requested to provide the following information for the internal events and internal flooding PRAs, as appropriate:

a. Conclusion 5 of NRC memorandum dated May 30, 2017 states, [t]he NRC staff does not agree with crediting spare portable equipment not modeled in the PRA in lieu of using appropriate failure rates, because this approach is not consistent with the ASME/ANS PRA Standard [ASME/ANS RA-Sa-2009] and RG 1.200. Furthermore, the potential impact of underestimating failure rates could be larger than the unquantified risk benefits of spare equipment not modeled in PRAs. Conclusion 6 of the memorandum states, [t]he failure rates of permanently installed equipment cannot be used for portable equipment even if sensitivity analyses are performed. Licensees should use plant-specific o[r] generic data collected and analyzed using acceptable approaches to estimate the failure rates for portable equipment.

of the LAR describes the FLEX strategies modeled in the PRA, including use of a portable diesel-driven generator (PDG) to restore power to specific equipment.

The failure rates for the PDG was developed using failure rates for an emergency diesel-driven generator (EDG) from NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants (2015 update)." The licensees basis for this assumption includes: (1) there are multiple spare portable generators on site that can be used if a single generator failed, and (2) the PDG failure rates will be considered as a source of uncertainty and a sensitivity study will be performed using the 5th and 95th percentile values. The basis for using EDG failure rates to represent that of the PDG is not consistent with Conclusions 5 and 6 of NRC memorandum dated May 30, 2017. To address the above observations, the licensee is requested to provide the following additional information:

i.

A detailed justification for using the EDG failure rates/probabilities to characterize the parameter estimates of the PDG and its uncertainties, and discuss how the PDG failure rates/probabilities are consistent with ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, and the NRC staff positions in memorandum dated May 30, 2017. Include in this discussion any plant-specific operational experience (e.g., number of failures, number of demands, operational hours) of the PDG, and discuss any screening or disregarding of plant-specific data (e.g., design modifications, changes in operating practices). Discuss how the failure rates/probabilities assumed in the PRA for the PDG is consistent with the relevant plant-specific evidence/operational experience.

OR ii. Alternatively to part (i), propose a mechanism to ensure that prior to implementation of the 10 CFR 50.69 risk categorization process the appropriate failure rates/probabilities for PDG (including common cause failures, as applicable) that meet the supporting requirements (SRs) in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, are incorporated into the PRA. Also, describe how these failure rates will be developed/estimated consistent with the applicable SRs under HLR-DA-C and HLR-DA-D in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200.

b. Section 7.5 of NEI 16-06 states that the maintenance procedures for the portable equipment should be reviewed for possible pre-initiator human failures that renders the equipment unavailable during an event. High-level requirement (HLR) HR-D in ASME/ANS RA-Sa-2009 describes the requirement assessing probabilities of the pre-initiator human failure events (HFEs) using a systematic process that addresses the plant-specific and activity-specific influences on human performance. Conclusion 13 of the NRC memorandum dated May 30, 2017 states, [u]ntil acceptable guidance is provided for identifying and assessing unique aspects of pre-initiator HFEs for mitigating strategies, the staff may request additional information regarding assessment of those human failure events.

of the LAR discusses the PRA modeling of FLEX operator actions as detailed in the emergency operating procedures. However, there is no discussion related to the process performed for the identification and assessment of the pre-initiator HFEs for mitigating strategies. To address the above observation, the licensee is requested to provide the following additional information:

i.

A detailed discussion on how pre-initiator HFEs for mitigating strategies were identified and assessed to meet HLR-HR-D of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200. This discussion should also address whether test and maintenance procedures for the portable equipment were reviewed for possible pre-initiator HFEs that renders the equipment unavailable during an event, and how the probabilities of those pre-initiator HFEs identified were assessed consistent with HLR-HR-D, as endorsed by RG 1.200.

ii. If pre-initiator HFEs for mitigating strategies were not assessed, then propose a mechanism to ensure that prior to implementation of the 10 CFR 50.69 risk categorization process the appropriate pre-initiator HFEs for mitigating strategies that meet the SRs in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, are incorporated into the PRA.

c. Condition (a) under SR HR-G3 of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, requires evaluation of the impact of the quality of operator training/experience when estimating human error probabilities (HEPs). Attachment 6 of the LAR describes the use of portable FLEX equipment for mitigating strategies. However, it is not clear to the NRC staff as to how operator training/experience was reflected in the HEPs associated with deployment and installation of the portable FLEX equipment (e.g., deployment could utilize non-trained personnel that do not have the same training given to operators). To address the above observation, the licensee is requested to provide the following additional information:

i.

Confirm whether or not non-trained operators (e.g., maintenance or security personnel) are used in the deployment and installation of the portable FLEX equipment.

ii. If non-trained operators are used in the deployment and installation of the portable FLEX equipment, then discuss how the impact of the quality of operator training/experience was evaluated in estimating the associated HEPs to meet SR HR-G3, as endorsed by RG 1.200.

RAI 04 - Dispositions of Key Sources of Uncertainty (APLA)

Paragraphs (c)(1)(i) and (ii) of 10 CFR 50.69 require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask importance of components. NEI 00-04 guidance states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, (ADAMS Accession No. ML17062A466), provides guidance for how to treat uncertainties associated with PRA models in risk-informed decisionmaking.

Additionally, Section 3.3.2 of RG 1.200, Revision 2, defines key assumptions and sources of uncertainty as follows:

A key assumption is one that is made in response to a key source of model uncertainty in the knowledge that a different reasonable alternative assumption would produce different results, or an assumption that results in an approximation made for modeling convenience in the knowledge that a more detailed model would produce different results. For the base PRA, the term different results refers to a change in the risk profile (e.g., total core damage frequency (CDF) and total large early release frequency (LERF), the set of initiating events and accident sequences that contribute most to CDF and to LERF) and the associated changes in insights derived from the changes in the risk profile. A reasonable alternative assumption is one that has broad acceptance within the technical community and for which the technical basis for consideration is at least as sound as that of the assumption being challenged.

A key source of uncertainty is one that is related to an issue in which there is no consensus approach or model and where the choice of approach or model is known to have an impact on the risk profile (e.g., total CDF and total LERF, the set of initiating events and accident sequences that contribute most to CDF and to LERF) such that it influences a decision being made using the PRA. Such an impact might occur, for example, by introducing a new functional accident sequence or a change to the overall CDF or LERF estimates significant enough to affect insights gained from the PRA.

As part of its audit (ADAMS Accession No. ML20058A010) of the licensees LAR dated December 6, 2019, the NRC staff reviewed the PRA assumptions and sources of uncertainty associated with the internal events and internal flooding PRA models. The dispositions of some PRA assumptions/sources of uncertainty were unclear to the NRC staff as to why they were not considered key for this application and evaluated under LAR Attachment 6. To address the above observation, the licensee is requested to provide the following additional information:

a. In order to confirm that the PRA assumptions/sources of uncertainty were properly assessed for this application, provide a more detailed justification for characterizing the following PRA assumptions/sources of uncertainty as not impacting the 10 CFR 50.69 application. If this justification considers the impact on plant risk (i.e., CDF and LERF),

address why this impact is an appropriate substitute for assessing SSC risk achievement worth (RAW) and Fussell-Vesely (F-V) impacts. If sensitivity studies were performed, describe these sensitivity studies and their results.

i.

A number of PRA assumptions/sources of uncertainty were dispositioned based on not impacting plant risk and that any potential impact will be handled when the system is categorized (i.e., identifier I10).

The NRC staff notes that minimal changes in plant risk can impact the categorization for some SSCs since RAW and F-V values, not plant risk values, determine the high-safety-significant (HSS)/LSS designations. Also, it is unclear to the NRC staff how any potential impact will be addressed during the categorization process.

ii. Environmental qualification (EQ) equipment in containment: Non-EQ equipment could be credited in the PRA. This may be a non-conservative bias for loss of coolant accident (LOCA) sequences where the licensee recommended a sensitivity study to determine its impact on certain applications.

iii. Direct current (DC) bus load shed failure during a station blackout (SBO) event:

SBO events can be a large contributor to risk. The conservative modeling decision of failure to load shed the DC bus during a SBO event could have a relatively large overall effect on risk, and therefore, the categorization results.

iv. The scenario cutoff time of two hours for the internal flooding analysis: The licensees characterization of this assumption recommended a sensitivity study.

However the associated disposition did not indicate a sensitivity study was performed nor was a basis presented for concluding that this assumption has no significant impact on the 10 CFR 50.69 SSC categorization.

v. Loss of offsite power (LOOP) recovery curves for human reliability analysis: The NRC staff observed that a sensitivity study was performed and determined the impact was not significant with regards to plant risk. However, minimal changes in plant risk can impact the categorization for some SSCs since RAW and F-V values, not plant risk values, determine the HSS/LSS designations.

OR

b. Alternatively, propose a mechanism to ensure that these PRA assumptions/sources of uncertainty will be appropriately addressed during the implementation of the 10 CFR 50.69 risk categorization process. Also, describe how these will be addressed during the 10 CFR 50.69 risk categorization.

RAI 05 - Interim PRA Updates (APLA)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant. RG 1.201, Revision 1, provides guidance for categorizations of SSCs.

Section 4.2 of RG 1.200 states the LAR should include a discussion of the resolution of the peer review facts and observations (F&Os) that are applicable to the parts of the PRA required for the application. This discussion should take the following forms:

A discussion of how the PRA model has been changed, and A justification in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application decision were not adversely impacted (remained the same) by the particular issue.

For F&O QU-F2-01 in Attachment 3 of the LAR, the peer review team identified where the licensees PRA update process does not address all aspects of SRs QU-F2 and QU-F3 in ASME/ANS RA-Sa-2009 when reviewing interim model updates. In the disposition for F&O QU-F2-01, the licensee explains that reverifying SRs QU-F2 and QU-F3 after interim PRA updates would not impact the application. It is not clear to the NRC staff how review of the new results is sufficient to conclude that repeating all the sensitivity studies related to the QU requirements is not expected to impact SSC categorization. The NRC staff notes improper truncation values that do not achieve convergence could potentially impact the 10 CFR 50.69 categorization results.

To address this observation, the licensee is requested to provide the following additional information:

a. A detailed justification that describes why not performing the truncation level sensitivity study for the PRA model used for 10 CFR 50.69 categorization does not impact the risk categorization of any SSC.

b. Alternatively, propose a mechanism to ensure that a truncation level sensitivity study is performed that confirms the truncation value(s) used to quantify the PRA does not impact SSC categorization prior to implementation.

RAI 06 - Categorization Process (APLA)

Paragraph (c)(1)(iv) of 10 CFR 50.69 requires that the SSC categorization process includes evaluations that provide reasonable assurance that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in CDF and LERF resulting from changes in treatment permitted by implementation of 50.69(b)(1) and (d)(2) are small. The guidance in NEI 00-04 states, [t]he purpose of the IDP [Integrated Decision-making Panel] is to ensure that the appropriate considerations from plant design and operating practices and experience are reflected in the categorization input. It is further discussed in Section 8.1 of NEI 00-04 that the cumulative sensitivity study should provide the IDP with both the overall assessment of the potential risk implications and the relative contribution of each system.

Lastly, Figures 1-2 and 2-1 of NEI 00-04 indicate that the risk sensitivity study should be provided to the IDP prior to SSC categorization.

Section 3.1.1 of the LAR states that [t]he order in which each of the elements of the categorization process (listed below) is completed is flexible, and, as long as they are all completed, they may even be performed in parallel. For certain elements, this does not appear to be the case (i.e., many of the elements must be completed prior to the IDP review). LAR Figure 3-1, Categorization Process Overview, depicts the cumulative risk sensitivity study following the IDP review. However, based on the above discussion, it seems this sensitivity study should appear before the IDP review in LAR Figure 3-1.

Thus, the licensee is requested to confirm whether or not the IDP review will consider information from the cumulative risk sensitivity study. If the IDP review will not consider this information, provide a detailed justification as to why the IDP review does not need to consider information from the cumulative risk sensitivity study, and how the requirements of 10 CFR 50.69 are met by the overall process depicted in LAR Figure 3-

1.

RAI 07 - SSCs Categorization Based on Other External Hazards (APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs. Section 3.2.4 of the LAR states that, all other external hazards were screened for applicability to SPS per a plant-specific evaluation in accordance with GL 88-20 and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the other external hazards screening results.

LAR Attachment 4, External Hazards Screening, states that Dominion Energy is in the process of reevaluating the external flooding hazard and the tornado missile hazard, and that any identified discrepancies will be tracked in the corrective action program. In light of these observations, the licensee is requested to address the following:

a. Provide a summary of the reevaluation of external flooding and tornado missile hazard including the basis for the reevaluation, potential differences between the current knowledge of those hazards at the site, and potential impact on SSC categorization under 10 CFR 50.69.

b. Discuss the licensees approach, such as an implementation item completed prior to implementation of the 10 CFR 50.69 program and controlled by the proposed license condition, to ensure that any impacts on SSC categorization under 10 CFR 50.69 identified after the completion of the updated external flood reevaluation and the updated tornado missile reevaluation are included in the program consistent with the guidance in NEI 00-04.

RAI 08 - Risk Contribution of a Seismic Event (APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific PRA, margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs. 10 CFR 50.69(b)(2)(iv) requires a description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv). The Statement of Consideration (SoC) on 50.69(b)(2)(iv) of the rule states that the licensee is required to include information about the evaluations they intend to conduct to provide reasonable confidence that the potential increase in risk would be small. The SoC further clarifies that a licensee must provide sufficient information to the NRC, describing the risk sensitivity study and other evaluations and the basis for their acceptability as appropriately representing the potential increase in risk from implementation of the requirements in the rule.

Section 3.2.3, Seismic Hazards, of the enclosure to the LAR states that the seismic risk (CDF/LERF) will be low such that seismic hazard risk is unlikely to influence an HSS decision.

Section 2.2.2 of the EPRI report 3002012988 identifies the contribution of seismic to total plant risk as a basis for the use of the proposed alternate seismic approach. Further, the insights in the EPRI report are derived from the full spectrum of the seismic hazard (i.e., the entire hazard curve). The NRC staff noted that the LAR does not provide sufficient information to support the claim that the plant-specific seismic risk is a small percentage contribution to the total plant risk such that an integral importance measure for a component would not result in an overall HSS determination, and thereby, the applicability of the proposed alternate seismic approach to the licensee.

The licensee is requested to provide a technical justification that supports the claim that the plant-specific seismic risk is low relative to the overall plant risk such that the categorization results will not be significantly impacted to support the applicability of the proposed alternate seismic approach.

RAI 09 - Configuration Control Process (APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs. Revision 3 of RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, (ADAMS Accession No. ML17317A256), establishes the need for an implementation and monitoring program to ensure that proposed changes do not degrade operational safety over time and that no adverse degradation occurs due to unanticipated degradation or common cause mechanisms. An implementation and monitoring program ensures that proposed changes continue to reflect the reliability and availability of impacted SSCs.

The staffs review of the LAR did not identify a description of the configuration control process to ensure that changes to the plant, including physical changes and changes to documents, are evaluated to determine the impact on design bases, licensing documents, programs, procedures, and training, or how the configuration control program is implemented.

The licensee is requested to provide a description of the configuration control process and a list for implementation of the configuration control program that includes:

a. A review of the impact on the System Categorization Document for configuration changes that may impact a categorized system under 10 CFR 50.69.

b. Steps to be performed if redundancy, diversity, or separation requirements are identified or affected. These steps include identifying any potential seismic interaction between added or modified components and new or existing safety-related or safe shutdown components or structures.

c. Review of impact to seismic loading, safe shutdown earthquake (SSE) seismic requirements, as well as the method of combining seismic components.

d. Review of seismic dynamic qualification of components if the configuration change adds, relocates, or alters Seismic Category I mechanical or electrical components.

RAI 10 - Alternate Seismic Approach (APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs. The proposed alternate seismic approach is based on insights of the EPRI report 3002012988, which derives risk insights from four case studies. Those case studies compare the HSS SSCs determined based on a seismic PRA (SPRA) against HSS SSCs determined from other PRAs used for categorization. Each of the cases studies included a full power internal events (FPIE) PRA, but only two of the four case studies used information from a fire PRA. Sections 3.3 through 3.5 of the EPRI report provide general information about the peer reviews conducted for the PRAs used for in each of the four case studies. However, the level of information is insufficient to determine whether the PRAs used in the case studies have been performed in a technically acceptable manner to support this application.

The NRC staff has previously requested and reviewed information to support its decision on the technical acceptability of the PRAs used in the case studies as well as details of the conduct of the case studies. This information is included in the supplements to the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, LAR for adoption of 10 CFR 50.69. The supplement to the 10 CFR 50.69 by Calvert Cliffs Nuclear Power Plant LAR, dated May 10, 2019 (ADAMS Accession No. ML19130A180), contained additional information related to the alternate seismic approach including incorporation by reference docketed information related to case study Plants A, C, and D; the supplement dated July 1, 2019 (ADAMS Accession No. ML19183A012) further clarified the information related to the alternate seismic approach (see response to RAI 4); the supplement dated July 19, 2019 (ADAMS Accession No. ML19200A216) provided responses to support the technical acceptability of the PRAs used for the Plant A, C, and D case studies as well as technical adequacy of certain details of the conduct of the case studies; the supplement dated August 15, 2019 (ADAMS Accession No. ML19217A143) clarified a response in the July 19, 2019 supplement. The supplement dated July 19, 2019 included modifications to the content of the EPRI report.

Since the above-mentioned information was requested and reviewed by the staff for Calvert Cliffs Nuclear Power Plants LAR for adoption of 10 CFR 50.69, the staff is unable to use it for the licensees docket unless it is incorporated in the licensees LAR. The above-mentioned information is necessary for the staff to make its regulatory finding on the licensees proposed alternate seismic approach and has not been provided by the licensee. Thus, the licensee is requested to address the following:

a. Provide the above-mentioned information to support the staffs regulatory finding on the alternate seismic approach by either incorporating the information by reference or responding to the RAIs in the identified supplements as well as providing information in the docketed documents related to case study Plants A, C, and D that were included by Calvert Cliffs Nuclear Power Plant in their supplement dated May 10, 2019 (ADAMS Accession No. ML19130A180).

b. If differences exist between the licensees proposed alternate seismic approach and the information in the supplement to the 10 CFR 50.69 by Calvert Cliffs Nuclear Power Plant LAR dated May 10, 2019 (ADAMS Accession No. ML19130A180), identify such differences and either incorporate them in the licensees proposed approach or justify their exclusion.

RAI 11 - Implementation Items (APLA/APLC)

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires that a licensees application contain a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown are adequate for the categorization of SSCs. If any of the responses to the RAIs associated with this LAR require follow-up actions prior to implementation of the 10 CFR 50.69 categorization process, the licensee is requested to provide a list of those actions and any PRA modeling changes including any items that will not be completed prior to issuing the amendment but must be completed prior to implementing the 10 CFR 50.69 categorization process.

Additionally, the licensee is requested to propose a mechanism that ensures these activities and changes will be completed and appropriately reviewed and any issues resolved prior to implementing the 10 CFR 50.69 categorization process (for example, a license condition that includes all applicable implementation items and a statement that they will be completed prior to implementation of the 10 CFR 50.69 categorization process).