ML20240A218

Request for Additional Information Regarding Amendment Request for Amendments to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Co

ML20240A218
Person / Time
Site:	LaSalle
Issue date:	09/03/2020
From:	Bhalchandra Vaidya Plant Licensing Branch III
To:	Bryan Hanson Exelon Nuclear
Vaidya B
References
EPID L-2020-LLA-0017
Download: ML20240A218 (15)

Text

September 3, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNIT NOS. 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO RENEWED FACILITY OPERATING LICENSES TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS." (EPID L-2020-LLA-0017)

Dear Mr. Hanson:

By letter dated January 31, 2019 Agencywide Documents Access and Management System (ADAMS) Accession No. ML20031E699), Exelon Generation Company, LLC (Exelon, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for an amendment to Renewed Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station, Units 1 and 2 (LaSalle).

Exelons proposed license amendment request (LAR) would modify the 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), Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors."

The NRC staff has reviewed your submittal and conducted a Regulatory Audit on June 15 to June 19, 2020. The NRC staff has identified areas where additional information is needed to complete its review. The draft request for additional information (RAI) is enclosed.

On August 18, 2020, the NRC staff held a teleconference with Exelon staff to clarify this request to ensure that the licensee clearly understood the NRC staff concerns. By email dated August 26, 2020, you committed to provide the responses to this request no later than September 30, 2020, for Question Nos. 1, 2, 8, 11, and 12, and October 22, 2020, for Question Nos. 3, 4, 5, 6, 7, 9, and 10.

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

Sincerely,

/RA/

Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-373 and 50-374

Enclosure:

Request for Additional Information cc: Listserv

REQUEST FOR ADDITIONAL INFORMATION LA SALLE COUNTY STATION, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS" Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, May 2006 Agencywide Documents Access and Management System (ADAMS) Accession No. ML061090627), endorses, with clarifications and qualifications, the Nuclear Energy Institute (NEI) guidance document NEI 00-04, Revision 0, 10 CFR 50.69 SSC [structures, systems, and components] for nuclear power reactors" Categorization Guideline, July 2005 (ADAMS Accession No. ML052910035), as one acceptable method for use in complying with the requirements in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69. The NEI 00-04 guidance describes in detail a process for determining the safety significance of SSCs and for categorizing them into the four risk-informed safety class (RISC) categories defined in 10 CFR 50.69. This categorization process uses an integrated decision-making process, incorporating both risk and traditional engineering insights. The NEI 00-04 guidance allows licensees to implement different approaches, depending on the scope of their probabilistic risk assessment (PRA). The proposed amendment includes an exception to the U.S. Nuclear Regulatory Commission (NRC) endorsed categorization process in NEI 00-04 to apply an alternative seismic approach specified in Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated July 2018[1].

The regulation 10 CFR 50.69(b)(2)(ii) provides the requirements to describe 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 other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs. The regulation 10 CFR 50.69(c)(ii) requires the categorization process to determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant-specific probabilistic risk analysis (PRA). The functions to be identified and considered include design bases functions and functions credited for mitigation and prevention of severe accidents. 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.

Finally, the regulation 10 CFR 50.69(e) requires periodic updates to the licensee's PRA and SSC categorization.

[1]

All references to the EPRI report in this document refer to EPRI 3002012988, Alternative Approaches for Addressing Seismic Risk in 10CFR50.69 Risk-Informed Categorization, July 2018. The same report is cited as Reference 4 in the enclosure to the licensees November 28, 2018 submittal and is publicly available free of cost online at https://www.epri.com/#/pages/product/000000003002012988/?lang=en-US.

Enclosure

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (ADAMS Accession No. ML090410014), describes an approach for determining whether the base PRA, in total or the parts that are used to support an application, is acceptable for use in regulatory decision making for light water-reactors. RG 1.200 endorses, with NRC staff clarifications and qualifications, the 2009 version of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard (ASME/ANS RA-Sa-2009).

Note 1: All references to the EPRI report in this document refer to EPRI 3002012988, Alternative Approaches for Addressing Seismic Risk in 10CFR50.69 Risk-Informed Categorization, July 2018. The same report is cited as Reference 4 in the enclosure to the licensees November 28, 2018 submittal and is publicly available free of cost online at https://www.epri.com/#/pages/product/000000003002012988/?lang=en-US.

Seismic Hazard Change and Performance Monitoring APLC 50-69 RAI 1:

Section 3.2.3 of the enclosure to the LAR cites the Moderate Seismic Hazard/Moderate Seismic Margin (so-called Tier 2 in the EPRI report) criterion from the EPRI report as being applicable to the licensee. The licensees criterion relies on the criteria for Tiers 1 and 3 in the EPRI report.

Section 3.2.3 of the enclosure to the LAR further states that In the unlikely event that the LaSalle [LSCS] seismic hazard changes from medium risk (i.e., Tier 2) at some future time, Exelon will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e). In addition, Section 3.5 of the enclosure to the LAR also states that implementation of the EDG [Exelon] design control and corrective action programs will ensure the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e) to performance monitoring as it pertains to the proposed alternate seismic approach.

It is unclear to the NRC staff how EGC will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) for the proposed alternate seismic approach.

Explain how the licensee will assess the need to update the SSC categorization and subsequently update, as appropriate, the SSC categorization for the proposed alternate seismic approach in accordance with 10 CFR 50.69(e) if the LSCS seismic hazard changes from medium risk (i.e., Tier 2) at some future time.

Application of the Proposed Alternate Seismic Approach APLC 50-69 RAI 2:

A review of the results of the case studies for Plants A, C, and D in Tables 3-5, 3-9, and 3-11 in the EPRI report, respectively demonstrate that the fragility groups identified as high safety significant (HSS) uniquely from seismic probabilistic risk analysis (SPRAs) include relay chatter induced failures and failures of passive components, NRC staff notes that a fragility group

includes multiple components. Section 3.6, Summary of Sensitivity Study Insights, of the EPRI report also discusses seismically-induced failures not limited to correlated and interaction failures.

Section 3.2.3 of the enclosure to the LAR as well as Section 4 of the EPRI report state that for the EDGs plant there may be a limited number of unique seismic insights appropriate for consideration in determining HSS, SSCs, and those SSCs would be typically associated with seismically correlated failures. The proposed approach includes a seismic walkdown of certain SSCs to look for correlation and spatial interaction concerns. Section 3.2.3 of the enclosure to the LAR also lists various seismic-specific failure modes that are included in the licensees review during categorization. Given the insights from the EPRI report and the expected seismic risk for the licensee, the basis for limiting additional evaluation in the proposed approach to only correlated and interaction issues is unclear.

Justify how unique seismic-specific failures other than correlated and interaction failures (i.e., those identified in the case studies and in Section 3.2.3 of the enclosure to the LAR) have been appropriately evaluated in the proposed approach without walkdowns and sensitivity calculations using the internal events PRA. Alternately, discuss how the proposed approach will evaluate unique seismic-specific failures identified in Section 3.2.3 of the enclosure to the LAR.

APLC 50-69 RAI 3:

Section 3.2.3 of the enclosure to the LAR includes a summary and flowchart (Figure 3-2 in the LAR) of the process to be followed for additional evaluations under the proposed approach. Additional information on the process is provided in Section 2.3.1 of the EPRI report. The discussion of the screening in Step 3 of Section 2.3.1 of the EPRI report states that [t]hese screening decisionslikely will involve cost/benefit decisions in terms of how best to complete the sensitivity study. Information on what cost/benefit decisions would be included, how such considerations would alter the implementation of the proposed approach, and the justification for such decisions as well as the resulting changes is unavailable either in the LAR or the EPRI report.

Step 3a of Section 2.3.1 of the EPRI report mentions inherently rugged components, cites Appendix H of Nuclear Energy Institute (NEI) 12-06, Revision 4 (ADAMS Accession No. ML16354B421), and lists certain SSCs as being inherently rugged. Appendix H of NEI 12-06, Revision 4, refers to other documents. It is unclear whether the list under step 3a of Section 2.3.1 of the EPRI report encompasses the inherently rugged components that are included in other documents referenced in Appendix H of NEI 12-06, Revision 4. The basis of consideration of such components as inherently rugged is also not provided.

Step 2 of Section 2.3.1 of the EPRI report discusses component classification and cites two different documents as resources for such classification. It is unclear to the NRC staff whether the classification(s) in the two cited documents consistent and the impact of any potential differences on the implementation of the proposed approach.

a. Discuss the justification for, intent of, and implementation of the consideration of the cost/benefit decisions in the screening process for the proposed alternate seismic approach. Include an explanation of how cost/benefit decisions would change the licensees proposed approach and justify such changes.

b. Provide the list of inherently rugged SSCs that will be used in Step 3a of Section 2.3.1 of the EPRI report. Include the basis of consideration of SSCs in the list as inherently rugged.

c. Provide the list of component classification groups that will be used in Step 2 of Section 2.3.1 of the EPRI report.

APLC 50-69 RAI 4:

Section 2.3.1 of the EPRI report provides the process for additional evaluation in the licensees proposed alternate seismic approach. Step 8 of the process states that the loss-of-offsite power (LOOP) and small loss-of-coolant accident (SLOCA) initiators in the Full Power Internal Event (FPIE) PRA will be used, in conjunction with surrogate events, to determine the impact of seismic-specific failures of SSCs following the walkdown. The basis for limiting the evaluation to only the LOOP and SLOCA initiators is not provided. It appears that the success path approach for seismic margins analysis (SMA) used in the Individual Plant Examination for External Events (IPEEE) is used to identify the initiators. However, the proposed approach is based on insights from seismic PRAs and not SMAs. Seismic PRAs include several initiators in addition to LOOP and SLOCA.

The initiating frequencies for the LOOP and SLOCA initiators in step 8 of Section 2.3.1 of the EPRI report are provided as 1.0 and 1E-2 per year, respectively. The assumptions used for the initiating frequency of SLOCA in the proposed approach are unclear and the basis for its generic applicability is not provided. The insights from the Plant A, C, and D, case studies that support the licensees proposed approach do not appear to inform the selection of the initiators and initiating frequencies. In addition, step 8a of Section 2.3.1 of the EPRI report states [o]ther appropriately justified values for small LOCA frequency may be used but does not provide any information about the basis for such values. Step 8c indicates that the sequence to be included in the evaluation is the SLOCA-LOOP sequence (i.e., SLOCA with conditional LOOP) rather than the SLOCA sequence.

The discussion in step 8 of Section 2.3.1 of the EPRI report does not provide any information about including changes to key internal events modeling assumptions due to seismic-specific impacts. Examples of such impacts include, but are not limited to, non-recovery of offsite power and non-recovery of direct current power. Such impacts contribute to the results from an SPRA and, therefore, are expected to be contributing factors to the insights from the plants A, C, and D, case studies in the EPRI report. It is unclear why such impacts are not being considered in the sensitivity study for the licensees proposed approach.

a. Discuss how other seismically induced failures (e.g., medium-sized (MLOCA),

modeling assumptions (e.g., non-recovery of LOOP), and all impacts of the seismically induced failure of the SSC being categorized were considered for inclusion in the FPIE SLOCA and LOOP event trees. If the sensitivity study excludes such failures and assumptions, justify their exclusion from the proposed approach.

b. Provide justification for the selected failure probability and the selected initiating frequencies for the LOOP as well as the SLOCA initiator using the insights from seismic PRAs, including the insights from the Plant A, C, and D, case studies used to develop the proposed approach.

c. Explain the basis for use of [o]ther appropriately justified values for small LOCA frequency initiating frequency SLOCA or SLOCA-LOOP sequences and how the proposed categorization process will ensure that the use of the other value is accepted by the NRC staff for the proposed alternate seismic approach (i.e., how the response to item (c) can be provided for such other value(s)).

The proposed approach utilizes walkdowns to identify correlated failure and interaction concerns and then use a sensitivity to determine their impact on the categorization of SSCs. The sensitivity calculation is a key step in the proposed alternate seismic approach and the resulting input to the categorization of SSCs. The proposed approach uses a failure probability for the surrogate events of 1E-4. Section 2.3.1 of the EPRI report states that the value is based a typical total seismic hazard exceedance frequency above which SPRAs would typically model loss of offsite power and for which correlated failures may be likely. The licensees proposed approach also states that the failure probability can be justified based on the hazard. Further, it appears that the same surrogate event failure probability is applied regardless of the SSC and seismic-specific failure mode that the event is supposed to represent.

d. Demonstrate that the proposed surrogate sensitivity study using a single failure probability for the surrogate event, the two initiating events, the corresponding initiating event frequencies, and any changes due to response to items (a) through (c) above, results in categorization input that is equivalent, or conservative compared to corresponding results from a SPRA. An example of such a demonstration is to use PRAs for one of the plants in the case studies in the EPRI report and compare categorization outcome from the SPRA for that plant and the surrogate sensitivity study using the FPIE for that plant. The NRC staff recognizes that the demonstration is not an actual categorization.

e. Explain how failure probability/ies other than a fixed value, determined based on the response to item (d), will be developed and justified based on the hazard for implementation in the proposed approach and how the use of the other value is accepted by the NRC staff for the proposed alternate seismic approach (i.e., how the response to item (e) can be provided for such other value(s)).

The proposed alternate seismic approach includes addition of surrogates to the FPIE PRA under the appropriate areas in the logic model (Step 7 of Section 2.3.1 of the EPRI report). The approach likens the surrogates to a common-cause failure mode. The FPIE PRA usually includes common-cause failure events under a particular random failure mode for an SSC (e.g., common-cause failure of pumps to run). However, the seismic correlated or interaction failures fail the SSC independent of a random failure mode of the SSC (i.e., the seismic failure is under the so-called top gate for the SSC). It is unclear where the surrogates will be included in the FPIE PRA.

The proposed alternate seismic approach compares the results of the sensitivity for each surrogate event to the results to the F-V and RAW HSS criteria for common cause failure in the FPIE PRA from NEI 00-04 (that is, F-V > 0.005 or RAW > 20). The proposed approach and the use of sensitivity for seismic-specific failure modes is based on the insights from the case studies in the EPRI report. It appears to the NRC staff that the objective of the sensitivity is to obtain categorization inputs like those that would be obtained from a seismic PRA for seismic-specific failure modes. However, the categorization using the seismic PRAs in the case studies in the EPRI report are not based on the common cause failure but rather the individual

SSC failure. As an example, Section 3.4.2.2 of the EPRI report, which discusses the identification of HSS SSCs from the Plant C SPRA, states that [c]omponents are considered HSS if the group F-V is greater than 0.005 or if the group RAW is greater than 2.0 for core damage frequency (CDF) or large early release frequency (LERF). Therefore, it appears that using different thresholds from those used for the case studies, which provide the insights supporting the proposed approach, is (i) not supported by the basis for the proposed approach, and (ii) can result in categorization inputs different those that would be obtained from a seismic PRA for seismic-specific failure modes, which is contrary to the purpose of the proposed approach.

f. Justify the use of the criteria for common cause failure for determination of HSS SSCs from the proposed sensitivity study given that the case studies providing the underlying insights for the proposed approach used different F-V and RAW criteria and that the proposed sensitivity study includes both correlated as well as interaction failure modes. Alternately, discuss any updated F-V and RAW criteria for use with the proposed sensitivity study and discuss their consistency with the case studies in the EPRI report as well as NEI 00-04.

APLC 50-69 RAI 5:

Step 10 in Section 2.3.1 of the EPRI report states that if the importance measure criteria are met by the surrogate basic events, then the corresponding SSC should be considered HSS. However, it is unclear whether the comparison against the importance measure criteria and the consequent categorization as HSS will be performed if the results from either of the sensitivities (i.e., for LOOP and for SLOCA) show the criteria being met or both the sensitivities need to show that the criteria is met or some combination shows the criteria is met. Further, the basis for the use of any approach (i.e., individual, combined, etc.) for comparison against the criteria is not provided.

Explain how the results from the sensitivity will be used for comparison against the importance measure criteria in NEI 00-04 (i.e., individually, combined, etc.). The explanation should include a justification of how the proposed comparison approach is consistent with the guidance in NEI 00-04 and reflective of the insights from the case studies in the EPRI report.

APLC 50-69 RAI 6:

Figure 2-3 in the EPRI report depicts the steps and decisions made as part of the licensees proposed alternate seismic approach. Certain steps in the flowchart in Figure 2-3 are unclear in their intent and scope. The accompanying discussion does not include information to address the lack of clarity in Figure 2-3.

Step 3b of Section 2.3.1 of the EPRI report states that if the SSCs under consideration is not used in safety functions that support mitigation of core damage or containment performance, it can be screened. The EPRI report also provides an example of screening based on this criterion. However, step 4 of Section 2.3.1 includes additional evaluation of SSCs where correlation and interaction considerations can impact the function of the SSCs. It is unclear whether the impact of the failure of SSCs that are screened out per step 3b of Section 2.3.1 of the EPRI report on SSCs with functions supporting mitigation is considered. For the example provided in the EPRI report, a chiller system that is not part of the safety function of the tank, the consideration of seismic-specific failure modes of the chiller system (e.g., correlated or interaction failure of the heat exchangers and or piping resulting in a drain path for the tank) can

change the screening decision. In addition, Section 4 of NEI 00-04 includes discussion of categorization of interfacing systems. It is unclear how the screening criterion in step 3b will interface with the guidance in Section 4 of NEI 00-04 for any interfacing systems.

a. Step 3b indicates a decision related to the SSC function for mitigation of core damage or containment performance. It is unclear if the mitigation function being questioned here are hazard-specific. Clarify whether the question in Step 3b includes SSCs functions for mitigation of core damage or containment performance for seismically-induced design basis and severe accidents. Include justification if the functions exclude mitigation of seismically-induced design basis and severe accidents.

b. The discussion for Step 3b in Section 2.3.1 of the EPRI report includes an example of a chiller system which would be screened out from consideration in the proposed alternate seismic approach. However, neither Step 3b nor the example in the discussion in Section 2.3.1 of the EPRI report considers the indirect impact of seismically-induced failures of SSCs on mitigation of core damage or containment performance. The case study for Plant C demonstrates such an impact. Table 3-9 of the EPRI report shows failures of two SSCs, one of them a chiller, (S_CB-CHLR-NCSW-FLOOD and S_1FC-ACU-FLD) which results in flooding and consequently loss of mitigation. Step 3b does not appear to include such impacts and would therefore, potentially screen out such SSCs. Clarify whether the question in Step 3b includes the indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance. Include justification if such impacts are excluded given the insights from the Plant C case study.

c. Step 5a through 5c in Figure 2-3 and the accompanying guidance in Appendices A and B of the EPRI report do not discuss the consideration and treatment of the indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance discussed in item (b) above. If indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance are included in Step 3b in response to item (b), explain how such SSCs will be considered in the walkdowns in Step 5a through 5c.

d. Explain how the consideration of indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance discussed in items (a) through (c) above interfaces with the categorization of interfacing SSCs discussed in Section 4 of NEI 00-04. If the proposed alternate seismic approach introduces changes to the approach in Section 4 of NEI 00-04, justify such changes.

APLC 50-69 RAI 7:

The steps in Section 2.3.1 of the EPRI report discuss the process for performing the sensitivity study. The surrogates used for the sensitivity study are intended to reflect the impact of seismic-specific failure modes identified in SPRAs. The steps in Section 2.3.1 of the EPRI report do not provide any guidance on whether the surrogates for SSCs that have already been categorized will be retained in the licensees FPIE PRA model for subsequent sensitivity studies (for other SSCs).

Explain, with justification, whether surrogates that were incorporated in the licensees FPIE PRA for the characterization of an SSC will be retained and included in the sensitivity studies for subsequent (and distinct) SSC categorizations.

APLC 50-69 RAI 8:

Appendix A of the EPRI report provides guidance on identifying seismic correlated or seismic interaction scenarios as part of the proposed alternate seismic approach. The guidance in that appendix includes statements indicating that the correlation decision is a judgment of engineers experienced in both seismic capacity and seismic response fields. In addition, steps for determining correlation provided in Appendix A of the EPRI report include consideration of similar seismic response, similar failure modes and fundamental frequencies, significantly different seismic responses, and interaction sources not deemed credible based on their experience and training. Such determinations appear to require technical knowledge and experience in areas such as seismic response, design, capacity, walkdown, and plant response. The requisite technical knowledge and experience is specialized and expected to be uncommon, especially with the plants staff. However, neither the LAR nor the EPRI report describes how the proposed approach ensures the qualifications of the personnel performing the walkdowns and using information related to failure modes, fundamental frequencies, plant response etc.

The walkdowns and result of the walkdowns appears to be an important element of the proposed alternate seismic approach. However, neither the LAR nor the EPRI report includes any information of how the walkdowns will be documentation and what the documentation will include. Such documentation appears to be necessary to support the conclusions of the walkdowns and for future regulatory processes, such as audits and inspections.

a. Provide the qualifications that personnel need to possess to perform the walkdowns and implementing the guidance in Appendix A of the EPRI report as part of the proposed alternate seismic approach. Discuss how the proposed approach ensures qualifications for performing the walkdowns and implementing the guidance. If the proposed approach does not explicitly consider such qualifications, justify not including special qualification considerations given that lack of specialized technical knowledge and experience indicated by the guidance in Appendix A of the EPRI report can decrease confidence in the results of the walkdowns.

b. Discuss how the approach ensures that the results of the walkdowns are communicated for incorporation in the proposed alternate seismic approach.

APLC 50-69 RAI 9:

Appendix B of the EPRI report provides guidance on capacity-based screening for high capacity SSCs. The guidance includes a screening value of the seismic core damage frequency and if a single SSC were to contribute that value or less, it could be screened from evaluation under the proposed alternate seismic approach. The proposed screening value appears to be an absolute value and therefore, not relative to the seismic risk for the licensee and the consideration of the importance measures thresholds used for categorization under 10 CFR 50.69. Depending on the seismic risk for the licensee, the proposed screening value can result in a non-trivial contribution relative to the seismic risk. Consequently, its use can result in screening of SSCs that would otherwise have to follow the other steps in Figure 3-2,

Seismic Correlated Failure Assessment for Tier 2 Plants, in the enclosure to the LAR (reproduced from Figure 2-3 in the EPRI report).

Appendix B of the EPRI report additionally provides guidance on the development of fragility values for SSCs to support the capacity-based screening. In addition to the state-of-practice approaches identified in Appendix B via relevant documents (e.g., representative values and conservative deterministic margins analysis [CDFM]), the guidance also discusses more simplified and conservative approaches when justified by experienced engineers, simplified approaches documented in ASCE 7, and assessments made would have to be necessarily conservative. In addition, Step 11 in Section 2.3.1 of the EPRI report discusses refinement to the fragility based on direction from the Integrated Decision-making Panel (IDP). SPRAs, such as those included in the case studies supporting the proposed alternate seismic approach, use state-of-practice approaches and undergo an NRC endorsed peer-review process for the implementation of such approaches. Further, based on reviews of several SPRAs that use the state-of-practice approaches, the NRC staff also has confidence in those analytical methods and their implementation by personnel with specialized knowledge about their use. The EPRI report and the LAR do not appear to include constraints on the type of approaches that can be used to achieve any refinement and therefore, such approaches can include those that have not been appropriately vetted even among the practitioners. It is unclear how the NRC staff can determine the acceptability of approaches that are not state-of-practice (either simplified or refined) and for which, as part of the proposed alternate seismic approach, an implementation peer-review will not be conducted.

It is expected that the fragility calculations require specialized knowledge and experience to implement the approaches as well as any caveats for the approaches in the corresponding documents. The qualifications of the personnel performing the fragility development, even using state-of-practice approaches, has not been specified in either the LAR or the EPRI report. Neither the LAR nor the EPRI report includes any information on the documentation of the fragility analysis. Such documentation appears to be necessary to support the conclusions of the analysis and for future regulatory processes, such as audits and inspections.

a. Justify the independence of the screening value of the seismic core damage frequency proposed in Appendix B of the EPRI report from the licensees seismic risk and the importance measures thresholds used for categorization under 10 CFR 50.69. Alternately, propose a screening approach relative to the licensees seismic risk which considers the impact on the importance measures thresholds used for categorization under 10 CFR 50.69.

b. Discuss how the calculation will be performed for comparison against the proposed screening threshold (or an updated threshold in response to item (a)). The discussion should indicate the seismic hazard curve (e.g., mean peak ground acceleration hazard curve) that will be used to perform the calculations.

c. Identify the analytical approaches and methods that are consistent with the state-of-practice and will be used for fragility calculations in the proposed alternate seismic approach.

d. Discuss how the proposed approach will ensure NRC staffs review and acceptance of analytical approaches and methods for fragility calculations other than those identified in item (c) above prior to their use in the licensees alternate seismic approach (e.g., through a LAR controlled by the license condition).

e. Discuss how the proposed approach will ensure that the qualifications of personnel that perform the fragility calculations, such as those using the methods identified in item (c), above are sufficient to support the development of the fragilities.

f. Discuss how the approach ensures that the fragility analyses performed for the proposed alternate seismic approach using the methods discussed in item (c) are documented to support the conclusions of the analyses and support future regulatory processes, such as audits and inspections.

APLC 50-69 RAI 10:

Table 3-1 of the enclosure to the LAR provides a categorization evaluation summary.

According to the table, the IDP can change HSS to LSS [low safety significant] for the seismic categorization step. Step 11 in Section 2.3.1 of the EPRI report that is incorporated by reference by the licensee (page 14 of the enclosure to the LAR includes Section 2.3.1 in the incorporation of the EPRI report in the LAR), states that the proposed approach (using the sensitivity study) is pseudo-deterministic and, therefore, the seismic correlated group HSS designations should be treated similar to HSS designations using the IPEEE SMA SSEL

[Safeguards Summary Event List] and in general, not be subjected to reconsideration by the Integrated Decision-Making Panel (IDP). It appears that the LAR is inconsistent with the EPRI report which it is referencing and supposedly, following. In addition, the basis for the LARs deviation from the EPRI report is also unclear.

Step 11 of the EPRI report states that SSCs which are HSS solely due to surrogate events representing seismic induced interactions (such as block walls impacting equipment) may be downgraded to LSS by the IDP with appropriate justification. The discussion on consideration of seismic correlated group HSS designations similar to IPEEE SMA SSEL and the subsequent possibility of downgrade to LSS is unclear to the NRC staff.

a. Clarify whether the summary for seismic categorization in Table 3-1 is intended to be a deviation from the approach discussed in the EPRI report. If yes, provide the justification for the deviation based on the insights and guidance in the EPRI report for the proposed approach which is incorporated by reference in the LAR.

If no, provide an updated version of Table 3-1 that clarifies the licensees intent.

b. Explain the apparent contradiction between two statements in the same step (step 11) in Section 2.3.1 of the EPRI report and clarify what is being proposed in terms of changes to HSS designations arising from the sensitivity study in the proposed approach by the IDP.

c. Explain, with examples, what would be appropriate justification by the IDP to downgrade an HSS determination arising from the sensitivity study in the proposed approach noting that a walkdown would have been conducted, the flowchart in Figure 2-3 of the EPRI report would have been followed, and a sensitivity would be performed before the HSS determination is made.

Technical Acceptability of PRAs Used for Case Studies in the EPRI Report APLC 50-69 RAI 11:

The proposed alternate seismic approach is based on the insights from the EPRI report which were derived from case studies. Those case studies compare the HSS SSCs determined based on a SPRA against HSS SSCs determined from other PRAs used for categorization.

Each of the cases studies included a 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 supporting this application have been performed in a technically acceptable manner.

The 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 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 August 15, 2019 (ADAMS Accession No. ML19217A143), clarified a response in the July 19, 2019 supplement (ADAMS Accession No. ML19200A216).

Since the above-mentioned information was requested and reviewed by the NRC 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.

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.

NON-SEISMIC EXTERNAL HAZARDS CONSIDERATION IN PROPOSED 10 CFR 50.69 PROGRAM APLC 50-69 RAI 12:

NEI 00-04 provides guidance on including external events in the categorization of each SSC to be categorized. Fire and seismic hazards are discussed in Section 5.2 and 5.3 of NEI 00-04, respectively. All other hazards are discussed in Section 5.4, Assessment of Other External Hazards. Figure 5-6 in Section 5.4 illustrates the process that begins with the SSC selected for categorization and then proceeds through the flow chart for each external hazard.

Section 3.2.4 of the enclosure to the LAR discusses the consideration of non-seismic external hazards in the proposed categorization approach. Attachment 4 of the enclosure to the LAR provides the basis for the consideration of each identified non-seismic external hazard in the proposed categorization approach.

The discussion for turbine missiles states, the speed capability of these rotors is considerably higher than the maximum attainable speed of these turbine generator units. Consequently, the probability of missiles being generated is statistically insignificant. The turbine missile probability analysis evaluates the failure of turbine stop, control, and bypass valves and determines the inspections and frequency of those inspection so that the failure rate and probability of turbine missile damaging safety related equipment is below the threshold of 10 7. It is unclear whether the turbine missile probability analysis is the basis for the screening or the speed capability of the rotors. It is also unclear whether turbine missile hazard will be subjected to the flowchart in Figure 5-6 to SSCs relevant to that hazard (e.g., turbine stop, control, and bypass valves).

The discussion for extreme winds states a demonstrably conservative estimate of CDF associated with high wind hazard (other than wind generated missiles) is much less than 1E-6/yr and [i]n addition, based on the plant design for tornado missiles, considering a limited set of SSCs vulnerable to tornado missiles, a demonstrably conservative estimate of CDF associated with tornado missiles is less than 1E-6/yr. Additional information to support the above cited statements is unavailable in the LAR.

Section 4 of Enclosure 4 to the licensees LAR for adoption of Technical Specification Task Force (TSTF)-505 (ADAMS Accession No. ML20035E577) includes information related to the impact of high winds on LaSalle. It is unclear if any or all the information from the licensees TSTF-505 LAR is applicable to the 10 CFR 50.69 LAR to support the demonstrably conservative estimates identified above. In addition, Section 4 of Enclosure 4 to the licensees TSTF-505 LAR includes discussion of operator actions and related equipment which can mitigate certain potential impacts from tornado-generated missiles. However, the licensees 10 CFR 50.69 LAR does not mention those actions and related equipment as being credited to support the consideration of high winds (aka extreme winds) in the proposed 10 CFR 50.69 categorization approach.

a. Identify the external hazards that will be evaluated according to the flow chart in Figure 5-6 of NEI 00-04.

b. Explain how any SSCs, including equipment required by applicable procedures, that are used to support the basis for screening high winds (aka extreme winds) and tornado-generated missiles in Attachment 4 of the enclosure to the 10 CFR 50.69 LAR will be considered in the proposed categorization approach and its consistency with the guidance in NEI 00-04 (e.g., Figure 5-6 in NEI 00-04 discusses SSCs that participate in a screened scenario).

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

ML20240A218 (*) By email OFFICE NRR/DORL/LPL3/PM(*) NRR/DORL/LPL3/LA(*) NRR/DRA/APLC(*)

NAME BVaidya SRohrer SRosenberg DATE 08/27/2020 08/27/2020 08/31/2020 OFFICE NRR/DORL/LPL3/BC(*) NRR/DORL/LPL3/PM(*)

NAME NSalgado BVaidya DATE 09/03/2020 09/03/2020