ML20218A642

R. E. Ginna Nuclear Power Plant - Request for Additional Information (RAI) for LAR to Implement WCAP, TSTF-411, and TSTF-418

ML20218A642
Person / Time
Site:	Ginna
Issue date:	08/05/2020
From:	V Sreenivas NRC/NRR/DORL/LPL1
To:	Tom Loomis Exelon Nuclear, Exelon Nuclear Generation Corp
Sreenivas V, 415-2597
References
EPID L-2020-LLA-0055
Download: ML20218A642 (8)

Text

From:

Sreenivas, V To:

Loomis, Thomas R:(GenCo-Nuc) (thomas.loomis@exeloncorp.com)

Cc:

Whitman, Jennifer; Danna, James; Dinh, Thinh; Biro, Mihaela; Dinsmore, Stephen

Subject:

R.E. GINNA NUCLEAR POWER PLANT: REQUEST FOR ADDITIONAL INFORMATION (RAI) FOR LAR TO IMPLEMENT WCAP, TSTF-411 AND TSTF-418, ( EPID: L-2020-LLA-0055)

Date:

Wednesday, August 5, 2020 1:28:00 PM REQUEST FOR ADDITIONAL INFORMATION (RAI) FOR LAR TO IMPLEMENT WCAP, TSTF-411 AND TSTF-418 ( EPID: L-2020-LLA-0055)

By letter dated March 25, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20085H900), Exelon Generation Company, LLC (the licensee) submitted a risk-informed license amendment request (LAR) to revise the R. E. Ginna Nuclear Power Plant (Ginna) By application, dated March 25, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20085H900),

Exelon Generation Company, LLC (the licensee) submitted a License Amendment Request (LAR) for the R.E. Ginna Nuclear Power Plant (Ginna). The proposed LAR would revise Technical Specification (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, and TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, based on Westinghouse topical reports WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, and WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times. The proposed changes are also consistent with the NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-411-A, Revision 1, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P) and TSTF-418-A, Revision 2, RPS and ESFAS Test Times and Completion Times (WCAP-14333), or are supported by plant-specific analysis for those changes which are plant specific, and therefore, not evaluated in these WCAPs.

The NRC staff has reviewed this licensees submittal and determined that additional information is required in order to complete the review. Please submit the responses to the following listed eight RAIs on the docket by September 11, 2020 to complete our review.

RAI 1 - PRA Model for the As-Built and As-Operated Plant

Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17317A256),

states that the engineering analyses conducted to justify the proposed licensing basis change should be based on the as-built and as-operated and maintained plant and reflect operating experience at the plant. The American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS-RA-Sa-2009 endorsed by RG 1.200 defines as-built, as-operated as a concept that reflects the degree to which the PRA matches the current plant design, plant procedures, and plant performance data, relative to a specific point in time.

Section 5.4.1, Plant Changes Not Yet Incorporated into the PRA Model in the License Amendment Request (LAR), states that the plant maintains an updating requirement evaluation (URE) database to track all enhancements, corrections, and unincorporated plant changes. It further states:

A review of all open URE items was performed for both Fire and [Full Power Internal Events] FPIE PRA models. In particular, a detailed review was performed on 57 High or Medium priority open UREs for the FPIE and/or Fire PRA models. No open items were identified that would have anything other than a negligible impact on the conclusions of [Technical Specifications Task Force] TSTF delta risk analysis or the TSTF results.

Describe the types of open items of high priority and explain how it was concluded that they have a negligible impact on the conclusions of TSTF delta risk analysis or the TSTF results.

RAI 2 - PRA Peer Review History

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014), provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RAS-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the Facts and Observations (F&Os) recorded by the peer review team and the subsequent resolution of these F&Os.

The ASME/ANS PRA standard RA-Sa-2009 defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of the ASME/ANS RA-Sa-2009 PRA standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this standard. Criteria presented to identify PRA upgrades are (1)use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

of the LAR describes the reviews conducted for the Ginna PRA. The internal events PRA was subject to a full-scope peer review in 2009 against ASME Standard RA-Sb-2005 and RG 1.200, Revision 1. The NRC staff is unclear, based on docketed information, whether there have been any upgrades to the internal events PRA that have not been peer reviewed. Address the following:

a) Summarize the model changes performed for the internal events, including internal flooding, PRA since 2009. This description should be of sufficient detail to determine whether the changes are considered PRA maintenance or PRA upgrades as defined in ASME/ANS RA-Sa-2009, Section 1-5.4, as qualified by RG1.200, Revision 2. For each change, indicate whether the change was PRA maintenance or a PRA upgrade, along with justification for this determination.

b) Confirm that focused-scope peer reviews have been conducted for any model change performed for the internal events, including internal flood, PRA model since July 2009

that meets the definition of a PRA upgrade, as defined in the ASME/ANS RA-Sa-2009 PRA standard. Describe the peer review and status of the resulting F&Os. Provide any remaining open F&Os, along with dispositions for this application.

RAI 3 - Disposition of PRA Facts and Observations (F&Os)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014), provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RAS-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os.

Internal events F&O IE-C10-01 found that the PRA documentation provided no explanation of differences between plant-specific initiating events and generic initiating events.

Disposition to F&O IE-C10-01 states that this issue is a documentation only issue.

However, in response to the request for additional information (RAI) 3 related to the Ginna Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) (ADAMS Accession No. ML16034A139), the licensee identified that there were differences with loss of bus initiating events which resulted an entry in the URE database. LAR Attachment 1 Table 5-2 states for URE 1202:

The subject electrical bus initiating events were reviewed for impact on this TSTF analysis. Any potential differences in Initiating Event frequencies would not have a significant impact on this analysis.

Discuss the differences in electrical bus initiating events and justify why it was concluded that they dont have impact on this application.

RAI 4 - Internal Flooding PRA

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014), provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RAS-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os.

LAR Attachment 1 Section 5.6.7 indicates that the increase in risk resulting from the changes proposed in this application is dominated by internal flooding sequences for both the internal events and the fire PRA:

the small increase in average [core damage frequency] CDF and [large early release frequency] LERF is mainly due to various internal flooding induced transient scenarios where [Auxiliary Feedwater] AFW fails to start automatically (one train is

failed by the flood), with the operators failing to start AFW manually []

A review of the Fire PRA cutsets shows a similar result. Various fire scenarios result in transients with a failure of AFW to start automatically (one train of signals is fire failed), followed by operators failing to start the pumps manually. []

The staff notes a number of F&Os were related to internal flooding. F&O IF-B2-01 (and similarly, F&Os IF-D6-01, IFEV-A7) identifies that the Ginna internal flooding PRA has a limited attempt to address human induced flooding mechanisms. F&O IF-D5a-01 identifies that the internal flooding PRA does not adequately address plant-specific characteristics that might affect the manner in which the frequencies of flooding are estimated.

Additionally, F&O IF-F3-01 identifies the lack of an adequate characterization of the sources of uncertainty associated with the flood analysis or a comprehensive discussion of the assumptions that could have an effect on the results.

LAR Section 5.8.4 attempts to address PRA key assumptions and sources of uncertainty but does not acknowledge the internal flooding contributors.

Address the following:

a) F&O IF-B2-01 (and similarly, F&Os IF-D6-01, IFEV-A7) identifies that the Ginna internal flooding PRA has a limited attempt to address human induced flood mechanism. In disposition to these F&Os the licensee states that discussion of human caused floods is discussed in detail in the internal flooding notebook, and that one maintenance induced flood was added to the model.

i. Provide a discussion of the systematic analysis that was performed to address human induced flooding mechanisms in the internal flooding PRA.

ii. Discuss whether this change was subject to a peer review, and if not, justify why not.

b) F&O IF-D5a-01 identifies that the internal flooding PRA does not adequately address plant-specific characteristics that might affect the manner in which the frequencies of flooding are estimated. In disposition the licensee states:

Regarding any effect on flood frequency due to aging affects, a sensitivity evaluation for a particular STI evaluation would show if there was any impact.

This does not appear to be a disposition applicable to this application. Provide an updated disposition for this F&O for the current application.

c) Describe the uncertainty evaluation performed for the internal flooding PRA to determine the assumptions and sources of uncertainty for the internal flooding. Provide any updated list of internal flooding key assumptions and sources of uncertainty and their associated disposition of impact on the application.

RAI 5 - Common Cause Modeling

According to Section A-1.3.2.1 of Appendix A of RG 1.177, when a component fails, the common cause failure (CCF) probability for the remaining redundant components should be increased to represent the conditional failure probability due to CCF of these components,

in order to account for the possibility that the first failure was caused by a CCF mechanism.

When a component fails, the calculation of the plant risk, assuming that there is no increase in CCF potential in the redundant components underestimates the calculated risk due to an entry in a TS limiting condition for operation, as illustrated by inclusion of the guidance in Appendix A of RG 1.177. Much of the discussion in Appendix A describes how configuration specific risk calculations should be performed.

TSTF-418, Section 4, provides the following guidance regarding plant-specific evaluations for functions not evaluated generically:

In order to apply the various relaxations justified in WCAP-10271 and WCAP-14333 to plant specific Functions not evaluated generically, a plant specific evaluation of those Functions and any additional plant specific Functions not listed in NUREG-1431 Rev. 1 but contained in the plant specific SSPS or RPS design must be performed.

The licensee performed a plant-specific analysis for its proposed CT extensions in LAR Section 5.6. LAR Attachment 1 Section 5.8 acknowledges the fact that the CCF could be a source of uncertainty. The disposition for this uncertainty only addresses the Reactor Trip System (RTS), stating:

For the [incremental conditional core damage probability] ICCDP/ [incremental conditional large early release probability] ICLERP calculations where selected components are set as failed, the approach conservatively adjusts the CCF failure probabilities for corresponding events for the RTS signal failure common cause event. This is considered conservative since not all failures would be subject to common cause failure modes. Therefore, this is not identified as a model uncertainty that could impact the decision.

With regards to Engineered Safety Feature Actuation System (ESFAS), LAR Attachment 1 Section 5.6.4.2.1 states:

Ginna does not model common cause for the ESFAS functions analyzed and therefore no changes are made to the model for the ICCDP/ICLERP calculation for the ESFAS functions. This I s considered acceptable since the WCAPs extensively analyzed the common cause failures modes that were critical to ESFAS and showed they were acceptable using a representative set of signals. Analyzing all of the ESFAS signals would result in a higher reliability of the signal portion of the risk analysis, since more signals would be available to actuate the system. Therefore, the WCAP risk analyses are considered bounding for common cause failures of ESFAS.

a. Explain what is meant by analyzing all of the ESFAS signals would result in a higher reliability of the signal portion of the risk analysis, since more signals would be available to actuate the system

b. Provide justification of why the referenced WCAP analyses are applicable and bounding of the Ginna plant specific analysis for common cause failures of ESFAS signals (related to containment pressure and steam line pressure) provided in this LAR or provide updated bounding estimates of risk due to the plant specific ESFAS completion time (CT) extensions proposed in this LAR.

RAI 6 - High Winds, External Flooding and Other External Events

TSTF-418, Section 4, provides the following guidance regarding plant-specific evaluations for functions not evaluated generically:

In order to apply the various relaxations justified in WCAP-10271 and WCAP-14333 to plant specific Functions not evaluated generically, a plant specific evaluation of those Functions and any additional plant specific Functions not listed in NUREG-1431 Rev. 1 but contained in the plant specific SSPS or RPS design must be performed.

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), states that the engineering analyses conducted to justify the proposed licensing basis change should be based on the as-built and as-operated and maintained plant and reflect operating experience at the plant.

The LAR provides an assessment of high winds, external floods and other external events from the Individual Plant Examination of External Events (IPEEE) study. LAR Section 5.4.5 states

High Winds, External Floods and Transportation Accidents were reviewed against the Standard Review Plan (SRP) []. Following plant modifications, it was determined that the Ginna plant met the Standard Review Plan criteria.

a) Since the IPEEE studies were performed in 1994 and have not been updated, discuss, in the context of the current plant and its environs, the applicability of the IPEEE conclusions for the current LAR.

b) In light of recent external flooding re-evaluation performed in response to the Fukushima Near Term Task Force (NTTF) recommendations, provide technical justification for why the risk from external flooding is negligible, or provide, with justification, a conservative or bounding estimate of the impact of external flooding risk for the current application.

RAI 7 - Plant Specific Risk Calculations

TSTF-418, Section 4, provides the following guidance regarding plant-specific evaluations for functions not evaluated generically:

In order to apply the various relaxations justified in WCAP-10271 and WCAP-14333 to plant specific Functions not evaluated generically, a plant specific evaluation of those Functions and any additional plant specific Functions not listed in NUREG-1431 Rev. 1 but contained in the plant specific SSPS or RPS design must be performed.

The licensee performed a plant-specific analysis for its proposed CT extensions in LAR Section 5.6. Tables 5-15 and 5-17 in LAR Section 5.5.6 present the ICCDP/ICLERP results for unavailability of the ESFAS/RTS instrumentation for internal events and respectively, for fire PRA. These tables show an ICCDP/ICLERP of 0.0 for the following entries: CONT-PRESS, OVR-TEMP, SG-WTR-LVL-A SG-WTR-LVL-B,

RCS-FLOW-A and RCS-FLOW-B. Additionally, Tables 5-14 and 5-16 in Section 5.5.6 present the delta CDF and delta LERF results for unavailability of the ESFAS/RTS instrumentation for internal events and respectively, for fire PRA.

a. Provide justification for the 0.0 values for ICCDP and ICLERP and explain why the risk is adequately captured for this configuration.

b. Describe and justify how the risk contributions from the signals discussed in item a.

above are captured in the delta CDF and delta LERF calculations.

c. Provide justifications for the 0.0 delta LERF presented in Table 5-16 for the fire PRA.

RAI 8 - Tier 3 Evaluations

RG 1.177, Revision 1, "An Approach for Plant-Specific, Risk-Informed Decision making:

Technical Specifications" (ADAMS Accession No. ML100910008), describes an acceptable risk-informed approach and additional acceptance guidelines geared toward the assessment of proposed permanent Technical Specifications (TS) CT changes. RG 1.177 identifies a three-tiered approach for the licensee's evaluation of the risk associated with a proposed TS CT change. Tier 3 addresses the licensee's overall configuration risk management program (CRMP) to ensure that adequate programs and procedures are in place for identifying risk-significant plant configurations resulting from maintenance or other operational activities and that the licensee takes appropriate compensatory measures to avoid risk-significant configurations that may not have been considered during the Tier 2 evaluation.

LAR Section 5.9.2, Tier 3. Risk-Informed Configuration Management, states:

Ginna uses the PARAGON Configuration Risk Monitor program []. For quantitative results, PARAGON links to the same fault trees and database as the internal events PRA model, so it is fully capable of evaluating CDF and LERF for internal events.

Address the following:

a) Explain and justify how the fire risk is addressed in the Ginna CRMP model.

b) When performing Tier 3 evaluations, explain whether the CRMP model at Ginna provides modeling of the reactor trip and ESFAS systems and components addressed by this LAR, including those addressed by WCAP-15376 and WCAP-14333.

If the CRMP model does not model relevant signals and components, please describe and justify how the CRMP evaluation is performed.

Docket Nos. 50-352 and 50-353

If you have any questions, please call me or email at v.sreenivas@nrc.gov V. Sreenivas, Licensing Project Manager Plant Licensing Branch I

Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation