ML20056D696

Request for Supporting Information for the Sequoyah SPRA Audit Review

ML20056D696
Person / Time
Site:	Sequoyah
Issue date:	02/19/2020
From:	Stephen Philpott Division of Operating Reactor Licensing
To:	Thompson R Tennessee Valley Authority
Philpott S
References
Download: ML20056D696 (3)

Text

From: Philpott, Stephen To: Thompson, Russell R Cc: Valentin-Olmeda, Milton

Subject:

Request for Supporting Information for the Sequoyah SPRA Audit Review Date: Wednesday, February 19, 2020 2:25:00 PM Good afternoon Russell, The purpose of this email is to request the following information to support the audit review of the Sequoyah Nuclear Plant (SQN) 50.54(f) seismic probabilistic risk assessment (SPRA) submittal dated October 18, 2019 (ADAMS Accession No. ML19291A003). The NRC staff is using a technical checklist (see ADAMS Accession No. ML17041A327) to guide this review. The following audit questions will support this effort. Please provide your responses to the following questions via your Certrec IMS Sequoyah SPRA audit site.

Fragility Clarification Questions:

Question 1 - Checklist Topic #4 - Adequacy of the Structural Model (SPID Section 6.3.1),

and Checklist Topic #8 - Screening by Capacity to Select SSCs for Seismic Fragility Analysis (SPID Section 6.4.3)

The SQN SPRA submittal discussed soil failure and fragility analysis in Section 3.3. Table 3.3-1 therein listed the structures with evaluated foundation materials concerning potential soil settlement and lateral deformation. Section 3.3.2 states that the resulting soil deformations were used as input to fragility analyses of important SSCs. The staff attempted to verify this statement for the Diesel Generator Building (DGB) and the Additional Diesel Generator Building (ADGB), both Category 1 structures. Based on the review of documents (1) CJC-SQN-C-001(Reference #46) and (2) TVAESQN010-REPT-001-Part A (Reference #36), the staff could not clearly identify how the estimated settlement of DGB and ADGB was used in the fragility analysis. Please clarify.

Question 2 - Checklist Topic #8 - Screening by Capacity to Select SSCs for Seismic Fragility Analysis (SPID Section 6.4.3)

The SQN SPRA submittal discussed the screening approach used in Section 4.4.1. The staff requests clarification for the following:

a. Table 4.4-1 shows that there are several screened out SSCs. Please clarify the basis for the screening of these SSCs from the SEL and its relation to the discussion in Section 4.4;

b. Based on the screening approach discussed in Section 4.4.1, it is not clear whether a capacity-based criterion was used for screening less rugged SSCs following the recommendation in SPID Section 6.4.3. Please clarify.

Question 3 - Checklist Topic #9 - Use of the CDFM/H Methodology for Fragility Analysis (SPID Section 6.4.1)

Appendix AE, Seismic Fragility Evaluation of the NSSS in Detailed Component Fragilities, for use in Second Risk Quantification of the Units 1 & 2 Sequoyah Nuclear Power Plant Seismic Probabilistic Risk Assessment, REPORT NO.: TVAESQN010-REPT-001 - PART B, Rev 1, states that The seismic fragility of the Nuclear Supply System (NSSS) is performed in accordance with the Hybrid Method The fragility of the NSSS components are based on the component specific design stresses, displacements, and seismic margins documented in FSAR (Ref. 3.6) and Westinghouse letters (Ref. 3.9 and 3.10). Wherever applicable, information from Watts Bar Nuclear (WBN) NSSS fragility documented in TVA calculation CDQ0000002015000710 (Ref. 3.7) are scaled appropriately to develop the fragility for SQN NSSS. Within the discussion of the Sub-Classes it was identified that the 1% damped SSE for Watts Bar was compared to the 5% damped RLE for SQN to develop scale factors (Appendix AE, page 11 of 48). Please provide clarification on how the differences in damping between the spectrum were accounted in the analysis.

Question 4 - Checklist Topic #9 - Use of the CDFM/H Methodology for Fragility Analysis (SPID Section 6.4.1)

Appendix B, Development of Structural Fragility for SQN SPRA in Representative Fragilities for Seismic Probabilistic Risk Assessment of Sequoyah Nuclear Power Plant, REPORT NO.: TVAESQN010-REPT-001 - PART A, Rev 2, states that Scaling method is implemented using the structural fragility of buildings at Watts Bar Nuclear Plant developed to support the SPRA, and modified by using the ratio between Ground Motion Response Spectra (GMRS) of the two sites at critical frequencies. It is recognized that the structures between the two sites are similar such that scaling of WBN structural fragilities can be used to identify SQN fragilities. Provide a clarification on the damping values adopted for the two sites, and if those values were different, how this difference was accounted for in the scaling process.

Plant-Response Model Questions Question 5 - Topic #16 - Review of Plant Modifications and Licensee Actions Sections 5.4 and 5.5 of the submittal provide CDF and LERF importance measures for risk-significant fragility groups and operator errors. It appears to NRC staff based on this information that there may be substantial cost-justified safety improvements that could significantly reduce the seismic CDF and/or LERF. Several SSC failures and operator errors, if eliminated, (individually or as part of a group) appear to have the potential to reduce the seismic CDF by 1E-05 per year or the seismic LERF by 1E-06 per year and be cost justified.

Therefore, please address the following:

a. Elimination or significant reduction in the probability of failure during a seismic event of each of the following fragility group failures and operator actions for Units 1 and 2 appears to have the potential to decrease seismic LERF by 1E-06 per year:

SEIS_0-30-5

HAMARV OP-LOCKOUT_69KSDB_S HACIV HAESBODG1_S HINST (for Unit 1 only)

Explain whether cost-justified plant improvement possibilities exist for Units 1 and 2 that would reduce the seismic LERF contribution by 1E-06 per year by eliminating or reducing the individual failure probability of these failures.

b. Elimination or significant reduction in the probability of failure during a seismic event of multiple operator actions for Unit 2 appears to have the potential to decrease seismic CDF by 1E-05 per year. These operator actions include:

OP-LOCKOUT_69KSDB_S OP-LOCKOUT_EDG_S HAMARV, HAESBODG1_S, HAFR2 Explain whether potential cost-justified plant improvements, such as by improving procedures or making equipment automatic, exist that would reduce the seismic CDF contribution for Unit 2 by 1E-05 per year associated with these operator actions (e.g., reduction or eliminating the failure probability for OP-LOCKOUT_69KSDB_S and OP-LOCKOUT_EDG_S). Include the rationale for any combinations that are not considered.

Please let me know when the responses are made available so that we can proceed with the audit review. If a conference call would be helpful to clarify or further explain any of these audit question, please let me know and I will be happy to arrange a call.

Thank you, Steve Steve Philpott Project Manager Nuclear Regulatory Commission (NRR/DORL/LPMB) phone: 301-415-2365 e-mail: Stephen.Philpott@nrc.gov