ML20248H388

Response to Request for Additional Information - License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System.

ML20248H388
Person / Time
Site:	Ginna
Issue date:	09/04/2020
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML20248H388 (43)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 September 4, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Response to Request for Additional Information - License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times

References:

1) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times, dated March 25, 2020 (ML20085H900)

2) Email from V. Sreenivas (U.S. Nuclear Regulatory Commission) to T.

Loomis (Exelon Generation Company, LLC), 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), dated August 5, 2020 (ML20218A642)

In the Reference 1 letter, Exelon Generation Company, LLC (EGC) requested changes to the Technical Specifications (TS) of the R. E. Ginna Nuclear Power Plant (Ginna). This proposed amendment revises TS 3.3.1, Reactor Trip System (RTS) Instrumentation, and TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation. These changes are 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.

In Reference 2, the U.S. Nuclear Regulatory Commission requested additional information.

Attached is our response to this request.

EGC has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in the Reference 1 letter. The supplemental information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a

U.S. Nuclear Regulatory Commission Response to Request for Additional Information - License Amendment Request for Implementation of WCAP-14333 and WCAP-15376 September 4, 2020 Page 2 significant hazards consideration under the standards set forth in 10 CFR 50.92. In addition, EGC has concluded that the information provided in this supplemental response does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments contained in this submittal. Should you have any questions concerning this submittal, please contact Tom Loomis at (610) 765-5510.

I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 4th day of September 2020.

Respectfully, David T. Gudger Senior Manager - Licensing Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information cc: NRC Regional Administrator, Region I NRC Senior Resident Inspector, Ginna NRC Project Manager, Ginna A. L. Peterson, NYSERDA

Attachment Response to Request for Additional Information

RISK MANAGEMENT TEAM Ginna ESFAS/RTS AOT Extension RAI Responses Notebook G1-LAR-006 Revision 1 September 2020

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses i

G1-LAR-006, Rev. 1 RM DOCUMENTATION NO.

G1-LAR-006 REV:

1 STATION:

Ginna UNIT(S) AFFECTED:

TITLE:

ESFAS/RTS AOT Extension - RAI Responses

SUMMARY

URE(s) Impacted:

None RM Document Level: Category 1, per ER-AA-600-1012.

[ X ] Review Required After Periodic Update

[ X ] Internal RM Documentation

[ ] External RM Documentation Electronic Calculation Data Files:

N/A Method of Review: [ X ] Detailed [ ] Alternate [ ] Review of External Document This RM documentation supersedes:

N/A in its entirety.

Prepared by:

Eric Thornsbury

/

9/1/2020 Name Signature Date Reviewed by:

Sam Falvo

/

9/1/2020 Name Signature Date Approved by:

Gene Kelly

/

Name Signature Date see email authorization at end of pdf

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses ii G1-LAR-006, Rev. 1 Previous Revisions:

REV.

DESCRIPTION PREPARER/DATE REVIEWER/DATE APPROVER/DATE 0

Original Issue Thornsbury 8/28/20 Falvo 8/28/20 Kelly 8/28/20 1

Revision based on TVT comments Thornsbury Falvo Kelly Configuration Control The Exelon Ginna PRA models are updated in accordance with ER-AA-600, Risk Management. This procedure requires periodic updates of the PRA Model and on-going reviews of design changes for PRA model impact. Model and Support Application documentation is controlled in accordance with ER-AA-600-1012, Risk Management Documentation. Any outstanding changes to the model or documentation are documented in the Ginna Updating Requirements Evaluations Database. This database is under the control of the Ginna Model Owner.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses iii G1-LAR-006, Rev. 1 TABLE OF CONTENTS Section Page 1.0 PURPOSE......................................................................................................... 1-1 2.0 RAI RESPONSES............................................................................................. 2-1 2.1 RAI 1 - PRA Model for the As-Built and As-Operated Plant................... 2-1 2.2 RAI 2 - PRA Peer Review History........................................................... 2-4 2.3 RAI 3 - Disposition of PRA Facts and Observations (F&Os).................. 2-8 2.4 RAI 4 - Internal Flooding PRA.............................................................. 2-10 2.5 RAI 5 - Common Cause Modeling........................................................ 2-19 2.6 RAI 6 - High Winds, External Flooding and Other External Events....... 2-22 2.7 RAI 7 - Plant Specific Risk Calculations............................................... 2-26 2.8 RAI 8 - Tier 3 Evaluations.................................................................... 2-30

3.0 REFERENCES

.................................................................................................. 3-1

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 1-1 G1-LAR-006, Rev. 1 1.0 PURPOSE The purpose of this document is to provide the basis for Risk Management related responses to NRC Requests for Information (RAIs) issued on August 5, 2020 [1] for the Ginna License Amendment Request (LAR) [2] to implement changes to Technical Specification (TS) requirements related to Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) instrumentation.

Consistent with the NRCs approach to Risk-Informed regulation, Ginna has identified particular TS requirements that are restrictive in nature and, if relaxed, have a minimal impact on the safety of the plant. These Technical Specifications require that various ESFAS and RTS instrumentation Allowed Outage Times (AOT) (also referred to as Completion Times [CT]) be restricted to a specific number of hours. The proposed changes are to increase specific ESFAS/RTS instrumentation AOTs from the currently specified time, based on WCAP-14333-P-A Rev 1 [3], WCAP-15376-P-A Rev 1 [4], TSTF-411 [5], TSTF-418 [6], and supplemental plant-specific analysis as documented in the LAR and supporting Risk Management documentation [7].

Each Risk Management related RAI question and response are documented in Section 2.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-1 G1-LAR-006, Rev. 1 2.0 RAI RESPONSES 2.1 RAI 1 - PRA MODEL FOR THE AS-BUILT AND AS-OPERATED PLANT REQUEST:

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.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-2 G1-LAR-006, Rev. 1 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.

RESPONSE

Table RAI-1 shows the description of each High priority URE and the basis for a negligible impact on the conclusions of this application. High priority is defined as items that could significantly impact applications or challenge the Exelon criteria for an unscheduled PRA update. Note that since the development of this application, some UREs have been downgraded and/or closed with no additional model changes that would impact the application.

TABLE RAI-1 REVIEW OF HIGH PRIORITY URES URE ID DESCRIPTION IMPACT ON TSTF-411/418 ANALYSIS 1121 Current modeling evaluates the Diesel Generator room doors as not watertight Due to the low expected change in risk from this evaluation and no direct interaction between this issue and RTS/ESFAS equipment, this conservatism will not mask risk insights from this evaluation.

This item is already closed in the Fire PRA model. This item has been recently downgraded to Medium priority.

1128 Various minor configuration risk management software issues This URE affects the configuration risk management tool, which is not used for this evaluation.

This item is already closed in the Fire PRA model. This item has been recently downgraded to Low priority.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-3 G1-LAR-006, Rev. 1 URE ID DESCRIPTION IMPACT ON TSTF-411/418 ANALYSIS 1179 F&O IF-C3-01 (IFSN-A6) includes discussion that Spray floods are not adequately documented in the IF Notebook.

Addressed in Table 5-3 of the LAR as part of the peer review findings. This URE documents that the Internal Flood Notebook needs Appendix C completed to complete documentation of spray impacts and modeling of additional spray floods if appropriate. The model used for this evaluation includes the necessary technical changes to address this URE. Therefore, there is no impact on the application.

This item is already closed in the Fire PRA model. This item has also been recently closed.

1202 Review electrical bus support system initiating event (SSIE) fault trees for Surveillance Test Interval (STI) change evaluations Addressed in Table 5-2 of the LAR as part of items identified in TSTF-425. Also see the response to RAI 3 below.

This item is already closed in the Fire PRA model.

1208 The new Installed Alternate Reactor Coolant System Injection Pump is in the PRA model; currently, for the Full Power Internal Events (FPIE) model, it is only credited for Station Black Out (SBO) Event Tree The current accident sequence modeling is a simplification. It would be a modeling improvement to better establish correlation between SBO Loss of Coolant Accident (LOCA) size and available water in Refueling Water Storage Tank.

This is a modeling improvement. The current modeling is acceptable but slightly conservative without including the new pump in all scenarios. Since the alternate pump has no tie to ESFAS/RTS, credit for the pump could slightly reduce risk calculations, including those when ESFAS signals fail, so current results are expected to be slightly conservative for this application.

This item is already closed in the Fire PRA model. This item has been recently downgraded to Low priority.

1237 Update PRA with installation of new Reactor Coolant Pump shutdown seals Although this URE remains open in the URE database, this change has been incorporated into the GN119A FPIE model used for this evaluation.

This item is already closed in the Fire PRA model. This item has also been recently closed.

1241 Instrument Air modeling error crediting long-term human actions under short-term IA gate IA000X Although this URE remains open in the URE database, this change has been incorporated into the GN119A FPIE model used for this evaluation.

This item is already closed in the Fire PRA model. This item has also been recently closed.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-4 G1-LAR-006, Rev. 1 URE ID DESCRIPTION IMPACT ON TSTF-411/418 ANALYSIS 1284 The design change will have three main portions: modification to the Standby Auxiliary Feed Water (SAFW) discharge valves 9701A & B (no FPIE impact), the SAFW isolation valves 9704A & B & 9746 (tornado valve)(no FPIE impact), and the SAFW recirculation AOVs 9710A & B (includes FPIE impact).

The FPIE impact of this change is minor and thus will not impact the results of this evaluation.

This item is already closed in the Fire PRA model. This item has been recently downgraded to Low priority.

2.2 RAI 2 - PRA PEER REVIEW HISTORY REQUEST:

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 RA S-2008, Standard for Level 1/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

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-5 G1-LAR-006, Rev. 1 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 RG 1.200, 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.

RESPONSE

a)

The Ginna full-power internal events (FPIE) model was peer reviewed [8] in 2009 and assessed against the ASME/ANS RA-Sa-2009 Standard [9]. In 2017 and 2020 the FPIE model has undergone Fact and Observation (F&O) close-out reviews [10][11] in accordance with NEI 05-04/07-12/12-06 Appendix X [12] and NEI 05-04 [13].

Improvement of the PRA model is an on-going and iterative activity. The model incorporates applicable design and procedure changes. Also, the model is continually

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-6 G1-LAR-006, Rev. 1 exercised, and the results reviewed to support applications such as performing on-line risk assessments or developing Mitigating System Performance Index analysis. As a result of these various activities, many incremental modeling changes are identified.

These modeling refinements and improvements include updating data in the model, fault tree structure changes, identifying mutually exclusive maintenance events, and removing conservatisms from flood initiating events and flood propagation changes. The modeling improvements and refinements are all done consistent with the methods used in the peer-reviewed model.

As part of the routine model update process, standard industry data sources (e.g. EPRI, NUREGs), along with site-specific data, are used to update data for initiating events, common-cause factors, and basic events in the model.

Many of the key changes in the Ginna FPIE model since the 2009 peer review were made to support the License Amendment Request for NFPA 805 [14]. Key model changes were reviewed by the NRC in subsequent audits, RAI responses, and the final safety evaluation

[15]. The plant changes are detailed in Attachment S of the NFPA 805 LAR. Key changes related to the PRA model include addition of new diesel generators, charging pump, tanks, upgraded reactor coolant pump seals, and portable equipment. The modeling changes include addition of related human actions based on updated emergency operating procedures. These human actions credit the equipment listed in Attachment S of the NFPA 805 LAR which includes portable equipment. This equipment is credited for any hazard the affects multiple redundant trains including, but not limited to, internal flood, station blackout, wind events, seismic events, fire events, and steam generator tube ruptures with isolation failure.

The NRCs NFPA 805 review did not include all changes to the FPIE model since 2009, as some are not relevant to the Fire PRA plant response model (such as internal flooding or steam-generator tube rupture related modeling) or occurred after the NFPA 805 review.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-7 G1-LAR-006, Rev. 1 Table RAI-2 identifies other key non-NFPA 805 changes in the Ginna FPIE model since the 2009 peer review, typically driven by design or procedure changes.

TABLE RAI-2 ADDITIONAL KEY MODEL CHANGES IN GINNA FPIE SINCE 2009 PEER REVIEW PRA MODEL CHANGE MAINTENANCE OR UPGRADE COMMENT Updated internal flood key operator actions Maintenance This is peer review F&O HR-G3-01, which was implemented in the GN119A model and determined to be Resolved in the 2020 F&O closure reviews. The review team concurred that this was a PRA Maintenance item.

Updated internal flood pipe whip/jet impingement and spray modeling Maintenance This is peer review F&O IF-C3-01, which was implemented in the GN119A model and determined to be Resolved in the 2020 F&O closure reviews. The review team concurred that this was a PRA Maintenance item.

Scrubbing credited for selected LERF scenarios Maintenance This is peer review F&O LE-C10-01, which was determined to be Resolved in the 2020 F&O closure reviews. Credit for scrubbing was not included in the model for this application. The review team concurred that this was a PRA Maintenance item.

Refined model by adding offsite power recovery basic events to the Level 2 model Maintenance Model improvement to use the same recovery basic events from the Level 1 model in the Level 2 model.

Refined flood propagation modeling for Air Handling Room and Battery Room floods Maintenance Changes consistent with the methodologies used in the peer-reviewed internal flood analysis. Additional analysis was performed including on how floods propagate through a door and through drain lines.

Updated modeling and basic event probabilities to used industry standard probabilities for LOOP recoveries Maintenance Change uses industry standard approaches.

Refined modeling of containment sump debris on long-term RCS cooling Maintenance Modeling consistent with industry standard approaches for PWR sump modeling.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-8 G1-LAR-006, Rev. 1 PRA MODEL CHANGE MAINTENANCE OR UPGRADE COMMENT Adoption of new tool for human action dependency analysis Maintenance Tool uses methodologies consistent with the previous analysis tool. Reviewed and found acceptable as part of 2012 FPRA ASME peer review and subsequent 2020 F&O closure review.

Updated flood frequencies to account for aging effects Maintenance Flood frequencies use standard industry data.

Incorporate portable and non-portable equipment NFPA-805 strategies and equipment and model changes, as allowed by updated operating procedures. Updated accident sequence modeling and human actions Maintenance Operating procedures updated based on PRA insights. Updated modeling and human action using existing methodologies in approved procedures.

b)

No PRA model changes since 2009 have met the definition of Upgrade to require a separate focused-scope peer review.

2.3 RAI 3 - DISPOSITION OF PRA FACTS AND OBSERVATIONS (F&OS)

REQUEST:

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 RA S-2008, Standard for Level 1/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

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-9 G1-LAR-006, Rev. 1 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.

RESPONSE

As noted in the RAI, 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), Ginna identified that there were differences with loss of bus initiating events which resulted in an entry in the URE database, listed as URE 1202. Furthermore, in supplemental information provided for TSTF-425 RAI 3 (ADAMS Accession No. ML16089A425), Ginna stated that Ginna PRA F&O IE-C10-01 has been resolved, and is now has (sic) a status of complete.

The technical portion of IE-C10-01 was Resolved in the 2017 F&O Closure [10] with model changes already incorporated in the GN119A model, and URE 1202 was initiated to complete the comparison of the electrical bus initiating event frequencies versus generic values since the peer review finding required a comparison of the plant calculated

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-10 G1-LAR-006, Rev. 1 value to the existing generic values, and an explanation of the differences. The URE states that the difference between the Ginna value and the generic value is the result of modeled operator recoveries, and that using the generic value would be a simplification.

Since the plant-specific value has a reasonable explanation and will continue to be used in the model, there is no impact to the TSTF-411/418 application.

2.4 RAI 4 - INTERNAL FLOODING PRA REQUEST:

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 RA S-2008, Standard for Level 1/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

[]

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-11 G1-LAR-006, Rev. 1 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.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-12 G1-LAR-006, Rev. 1 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.

RESPONSE

a) i)

Note that F&Os IF-B2-01 and IF-D6-01 are listed as Complete in LAR Table 5-3 and were Resolved in the 2017 F&O Closure [10]. The closure process concluded that the approach taken to consider the impacts of human induced flooding was adequate and appropriate for IF-B2-01. For IF-D6-01, the closure report provides a summary of the systematic analysis that was performed: A more thorough search was made for floods that could be induced by human action or other failures during maintenance activities. The types of isolation measures taken for fluid systems were examined, and this examination was supplemented by an examination of plant-specific operating experience and interviews with plant personnel. Only one source of flooding -

that associated with hardware failure of an isolation valve during major maintenance on a service water heat exchanger was identified for explicit modeling.

ii)

The resolution of F&Os IF-B2-01 and IF-D6-01 did not result in the application of new methods to the PRA model and did not have a significant impact on

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-13 G1-LAR-006, Rev. 1 significant accident sequences, so their resolution is categorized as model maintenance and no focused scope peer review was deemed necessary. The 2017 F&O Closure concurred with the designation as Maintenance.

b)

Note that F&O IF-D5a-01 has been Resolved in the 2017 F&O Closure [10]. The first portion of the disposition of F&O IF-D5a-01 is correct in that plant-specific issues including water hammer and human-induced floods have been addressed in the Internal Flooding PRA model and documentation. Pipe break frequencies are based on the appropriate age range of the plant per the EPRI methodology. Aging effects were included in the GN119A model per URE 1153, so no further sensitivity is required.

Because this F&O was Resolved for the model used for this application, there is no impact on this application.

c)

To assist with a determination of the impacts of internal flooding uncertainties on this application, Table RAI-4-1 shows the internal flood events that had the largest increases in Fussell-Vesely importance from the base case to the ESFAS-RTS-UA case with assumed unavailability increases. Only the global spray scenario event listed is above the typical 0.005 threshold for risk-significance, and no other risk-significant flood showed an increase of any amount. Flood Area IBN-1 consists of the north section of the Intermediate Building at elevation 253'-6". It contains the auxiliary feedwater pumps, various primary system instrumentation and control rod drive cabinets and MG sets.

Table RAI-4-2 lists all of the existing assumptions and sources of uncertainty identified in the Internal Flood documentation [16]. This list is exhaustive and not limited to only the key uncertainties. No new assumptions or sources of uncertainty were identified unique to this application. A disposition of each source of internal flood uncertainty for this application is listed in Table RAI-4-2 with a focus on the risk-significant floods identified in Table RAI-4-1.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-14 G1-LAR-006, Rev. 1 TABLE RAI-4-1 FLOODS WITH GREATEST FV CHANGE BASIC EVENT FLOOD EVENT BASE FV FV CHANGE FL-IBN-GSPR Global Spray Scenario in IBN-1 0.02023 0.00321 FL-IBN-SW-B-2K 2000 gpm IB SW Header B Flood 0.00124 0.00013 FL-IBN-SW-D-2K 2000 gpm IB SW Discharge Header Flood 0.00123 0.00012 TABLE RAI-4-2 REVIEW OF FLOOD UNCERTAINTIES SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS Large Turbine Building Floods. Assumed that all Turbine Building flooding would be contained on the 253' elevations of the Turbine Building and the hallways of the Service Building directly open to the rollup door separating the Turbine Building and the Service Building.

Large Turbine Building Floods were not a significant contributor to the risk changes in this application per Table RAI-4-1. No impact on this application.

Battery Room and EDG Doors. Assumed that components in these rooms will experience failure when the flood level in the Turbine Building is to a level of critical components in these rooms. This is conservative, since these doors may limit the water inflow to a rate that could be accommodated by the sump and drains in these rooms.

Conservative assumptions about Battery Room and EDG doors were not a significant contributor to the risk changes in this application per Table RAI-4-1. No impact on this application.

Intermediate Building Drains. Assumed that a 15,000 gpm flood in this area would exceed the capacity of these manholes and result in equipment failure in this area prior to the sub-basement being filled.

Conservative assumption about Intermediate Building drains was not a significant contributor to the risk changes in this application since the drain size has no impact on the spray scenario in IBN-

1. Larger Intermediate Building floods are not a significant contributor per Table RAI-4-1.

SW breaks of 100 gpm or less will not impact cooling to large SW loads.

Small SW breaks (i.e., <100 gpm), such as a spray scenario in IBN-1, would not be expected to impact cooling to large SW loads as this uncertainty assumes. Since this is a reasonable assumption, it is not considered to significantly impact the risk changes in this application.

Turbine Drive AFW and Motor Driven AFW pumps are assumed to fail in two hours without SW Cooling.

For the spray scenario in IBN-1, SW cooling is assumed to continue per the previous assumption. If either pump were failed due to the spray, this assumption would have no impact.

Therefore, it is not considered to significantly impact the risk changes in this application.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-15 G1-LAR-006, Rev. 1 SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS If Safety Injection is required, it is assumed that switchover to high pressure recirculation will be required in 100 minutes. This is based on the minimum time to deplete the RWST, and assumes operation of both safety injection and containment spray. This would be conservative for many LOCA scenarios.

The top cutsets for the spray scenario do not include operator actions using this assumption, which is also conservative. Any improvement in the modeling of the action would reduce the impact, so it is not considered to significantly impact the risk changes in this application.

Spray from piping will be directional and the impact will typically be quite local. Assumed that any spray event occurring within 16 feet of the identified target would impact that target. This is conservative, because there would be high probability that the resulting spray would be in a direction that would not threaten the identified target.

This assumption does impact the risk-significant spray scenario, but does so in a conservative manner. Any improvement in the modeling of the action would reduce the impact of the risk changes proposed in the LAR.

Pipe break frequencies were grouped by floods of less than 100 gpm (spray), 100 gpm to 2,000 gpm (flood), and 2,000 gpm to the capacity of the system (major flood). For each of these categories, the break flow was assumed to be the maximum for that category, unless system characteristics (e.g., pipe size, pressure, etc.)

limited the maximum flood rate achievable. This is somewhat conservative because with the group, a spectrum of break sizes is possible. Also, the pressure used to calculate maximum flow rates was typical of the discharge pressure of the pumps. No reduction of pressure was assumed for break locations far downstream of the pumps.

These factors result in some conservatism in the calculated break flow rates.

Conservatisms in grouping floods per this assumption would not impact the risk-significant flood. Therefore, it is not considered to significantly impact the risk changes in this application.

Damage to electrical cabinets was assumed to occur when the flood level reached the bottom of the cabinet. Cabinet interiors were not examined to determine the height of vulnerable components within.

Since spray events only damage by direct impact and not accumulation of water, this is not considered to significantly impact the risk changes in this application.

Operability of equipment in beyond design basis environments. Due to the scope of PRAs, scenarios may arise where equipment is exposed to beyond design basis environments (e.g., without room cooling, without component cooling, deadheading of low-pressure pumps in small LOCA scenarios, etc.).

Credit for operation of systems beyond their design-basis environment is not taken. Therefore, it is not considered to significantly impact the risk changes in this application.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-16 G1-LAR-006, Rev. 1 SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS Piping failure mode. The frequency of floods of various magnitudes (e.g., small, large, catastrophic) from various sources (e.g., clean water, untreated water, salt water, etc.) is an important input to the internal flooding analysis.

The internal flood analysis and initiating event frequencies for spray, flood, and major flood scenarios are developed consistent with the EPRI methodology. The use of generic flood frequencies with plant-specific estimates of pipe lengths is suitable for representation of the flood frequencies at the site. This is considered an accepted industry practice, so is not considered to significantly impact the risk changes in this application.

Basis for HEPs. Human failure events are typically significant contributors to CDF.

System or accident sequence modeling incorporates Human Failure Events (HFEs) and their associated Human Error Probabilities (HEPs). HEP estimation is based on accepted methods used for all HFEs, so is not considered to significantly impact the risk changes in this application.

Treatment of HFE dependencies.

HFEs associated with internal flood HEPs have been explicitly modeled as being independent of other HFEs. Top cutsets for the IBN-1 scenario do not include HFE dependencies, so this is not considered to significantly impact the risk changes in this application.

Treatment of rare and extremely rare events.

Selection of data should be based on confirmation that the database used is applicable to the plant.

Data for flooding frequencies are developed consistent with EPRI data and methodology. The use of generic flood frequencies with plant-specific estimates of pipe lengths is suitable for representation of the flood frequencies at the site.

This is considered an acceptable industry approach, so is not considered to significantly impact the risk changes in this application.

CDF and LERF definitions.

The definition of CDF and LERF is consistent with the internal events CDF and LERF definition and based on industry good practice. Therefore, it is not considered to significantly impact the risk changes in this application.

Engineering analyses - separate engineering analyses may use codes or invoke other assumptions that may introduce potential sources of modeling uncertainty.

Codes and methods applied in the Ginna internal flood PRA model are either accepted by the NRC or are considered good engineering practices, so is not considered to significantly impact the risk changes in this application.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-17 G1-LAR-006, Rev. 1 SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS Passive system degradation mechanisms - aging of active components is incorporated into the periodic data analysis updates but passive system reliability is generally not accounted for.

Data for flooding frequencies are developed consistent with EPRI data and methodology. The use of generic flood frequencies with plant-specific estimates of pipe lengths is suitable for representation of the flood frequencies at the site.

This is considered an acceptable industry approach, so is not considered to significantly impact the risk changes in this application.

Water hammer impacts on system performance.

Water hammer events are typically one of the causes for rupture of a water system, which is inherently included in the evaluation of the pipe rupture rates. High Energy Line Breaks are considered within the plant design basis. This is not a significant contributor to the risk changes in this application since water hammer is not applicable to the spray scenario in IBN-1. No impact on this application.

Plant partitioning. Assumptions made for a plant partitioning approach could be a source of model uncertainty.

Plant partitioning has been done considering the actual layout of the rooms in the plant. Each flood zone analyzed in the Ginna internal flood PRA model corresponds to a physical room in the plant. Partitioning does not impact the spray scenario in IBN-1, so is not considered to significantly impact the risk changes in this application.

Spray flood scenarios. Spray flood scenarios with less than 100 gpm flow do not totally disable the system with which they are associated. Also, spray effects consider only those components within a ten-foot radius (3 meters) and in the line-of-sight of a pressurized source.

Spray initiator scenario impacts are limited to the local effects of the spray. This assumption is realistic since such a low flowrate would not affect most systems needed to mitigate an accident since they have much larger flowrates.

Also, to account for any uncertainty in length estimation and to provide a more thorough accounting of potential spray sources, a radius of influence of 16 feet (5 meters) was used in the analysis. This is consistent with the guidance provided, so is not considered to significantly impact the risk changes in this application.

Flood and Major Flood sources are assumed to totally disable the system with which they are associated.

Floods and Major Floods were not a significant contributor to the risk changes in this application per Table RAI-4-1. No impact on this application.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-18 G1-LAR-006, Rev. 1 SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS Adequate drainage.

In general, floor drains were not credited in mitigating the larger flooding events, but were credited in mitigating flooding events with flow rates that did not exceed the installed drain capacity. Propagation of flooding due to conveyance of water via floor drains was considered for propagation of water to other areas of the plant. Drainage is not a significant contributor to the risk changes in this application since the drain size is not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

Walls and floors. Walls and floors provide physical separation between flood areas/zones. Failure of walls and floors may generate different propagation paths.

Walls and floors have sufficient structural and leak tight capability to preclude flood propagation beyond these areas. They are not a significant contributor to the risk changes in this application since they are not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

Fire doors. Door failure may lead to different flood propagation pathways.

In general, non-watertight doors were not credited in limiting the propagation of flooding events. They are not a significant contributor to the risk changes in this application since they are not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

Watertight doors. Door failure may lead to different flood propagation pathways.

The only watertight doors that are credited in the model are those protecting the EDG rooms. They are not a significant contributor to the risk changes in this application since they are not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

Hatches. Hatch failure may lead to different flood propagation pathways.

Catastrophic failure of floor hatches was not considered a credible failure mechanism due to the head of water being insufficient to challenge their structural integrity, except for only the largest of flooding events with flow rates capable of overwhelming drainage flowpaths. This is not a significant contributor to the risk changes in this application since it is not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-19 G1-LAR-006, Rev. 1 SOURCE OF UNCERTAINTY FOR INTERNAL FLOOD IMPACT ON TSTF-411/418 ANALYSIS HVAC Ducts. HVAC Ducts can be flood propagation pathways depending on their location.

HVAC ductwork is not generally assumed to be watertight. Ventilation openings are also noted as being located in the overhead elevations for most rooms, thus not affording a propagation pathway.

This is not a significant contributor to the risk changes in this application since it is not applicable to the risk-significant spray scenario in IBN-1. No impact on this application.

2.5 RAI 5 - COMMON CAUSE MODELING REQUEST:

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.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-20 G1-LAR-006, Rev. 1 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 is 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

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-21 G1-LAR-006, Rev. 1 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.

RESPONSE

a)

The phrase 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 is meant to reflect the fact that more than one ESFAS signal is typically available to actuate the system for any given scenario. PRA models, including those utilized in the WCAP, typically analyze only a subset of the potential signals. Addition of other diverse, redundant ESFAS signals to the PRA would be expected to decrease the importance of any individual signal.

b)

The WCAP analysis of ESFAS systems addressed a sampling of representative ESFAS systems (e.g., see Section 3.0 of WCAP-14333 [3]). While the specific systems do not necessarily match those analyzed in the Ginna plant-specific analysis supporting this application, the same impacts of common cause are expected to apply since different ESFAS systems have similar structures with multiple channels of signals feeding through similar components (i.e., bistables, relays, etc.). In accordance with the response to part (a), since multiple diverse/redundant signals are expected to actuate ESFAS, but only a subset of the expected signals are actually included in PRA models, failures within one set of inputs (including due to common cause) are not expected to have significant impacts. The analysis in the WCAP of ESFAS systems confirms this understanding.

However, to provide additional insight during development of the risk calculations for this application, for the modeling of ESFAS signals related to containment pressure and steam line pressure (items 10 and 11 in LAR Section 5.6.4.2), new events were inserted under all logical gates that represent failure of the SI signal to actuate the relays at the

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-22 G1-LAR-006, Rev. 1 auxiliary relay level. One event addresses all Train A ESFAS actuations, and one event addresses all Train B ESFAS actuations. Note that the specific gates do not model individual signal failures, so the inserted Train A/B event conservatively covers all failure modes including common cause within or among different signals and channels.

The Ginna PRA model does include common cause failures of components such as master relays, but these are at a lower level of logic than the inserted events for this application, so modeling failure of the inserted events is more conservative than adjusting such lower-level common cause terms. The results shown in the LAR incorporate this modeling and meet the risk thresholds.

Therefore, while the WCAP analysis may be deemed sufficient to address various common cause combinations of ESFAS failures, it is not solely relied upon to justify this application. To further support the estimate of risk due to this application, the conservative modeling implemented during the plant-specific calculations also addresses failures within or among signals due to common cause and provides a bounding estimate of risk.

2.6 RAI 6 - HIGH WINDS, EXTERNAL FLOODING AND OTHER EXTERNAL EVENTS REQUEST:

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.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-23 G1-LAR-006, Rev. 1 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.

RESPONSE

a)

As noted in Section 5.4.5 of the LAR, the primary external events of concern identified by the IPEEE were internal fire and seismic risk. Internal fire events are now modeled by the Fire PRA, and the IPEEE is not relied upon for that hazard. In the context of the current plant and its environs, there are no significant changes to the plant or environs; however, additional analyses have been performed since the IPEEE to provide

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-24 G1-LAR-006, Rev. 1 updated assessments of various hazards. Additional evaluations of external hazards regarding seismic and high winds events are discussed below. In addition, Section 5.1.3 and 5.8.6 of the LAR note that all other external hazards were screened out as being insignificant, though external flooding was given additional consideration in response to the Fukushima Near Term Task Force (NTTF) recommendations, discussed in part (b),

and seismic risk was further considered in Section 5.6.6.1 of the LAR. Also noted in Section 5.4.5, high winds and transportation accidents were also reviewed against the Standard Review Plan and found to be acceptably low. Plant changes such as the addition of FLEX equipment generally help with response to external events and do not create new vulnerabilities. IPEEE conclusions are referenced only for historical insights and are not used directly for fire or seismic analysis relative to this application.

Regarding seismic risk, in addition to the discussion in Section 5.6.6.1 of the LAR, note that the closeout of the IPEEE resulted in a minimum of a 0.2g Review Level Earthquake for Ginna. Subsequently, as part of NTTF Recommendation 2.1 Seismic, a site-specific assessment was made between the Ginna Safe Shutdown Earthquake anchored at 0.2g and the Ground Motion Response Spectra (GMRS) calculated by the NRC. In an NRC SER dated February 18, 2016 [17], the NRC acknowledged that the SSE was greater (at most frequencies much greater) than the GMRS, demonstrating robust seismic margin.

Only at frequencies between about 30Hz and 40Hz [18] was there a slight exceedance of the Safe Shutdown Earthquake, which the NRC determined to be of no concern, and no additional issues were identified for Ginna Station relative to seismic criteria. The combination of this review for NTTF and the analysis in Section 5.6.6.1 of the LAR support the conclusion of minimal risk impact from seismic events on this application without specific reliance on the IPEEE.

Regarding high winds, tornadoes, and missiles, aside from the IPEEE, Ginna also committed to include tornadoes and tornado missiles into its licensing basis as part of the Systematic Evaluation Program [19] performed to convert Ginnas Provisional Operating License to a Full-Term Operating License. The commitment was to protect the Reactor Coolant Pressure Boundary, the spent fuel assemblies, one train of equipment required

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-25 G1-LAR-006, Rev. 1 to achieve safe shutdown, and to prevent accidents that could result in radiological releases greater than Part 100 limits. The characteristics of this licensing basis tornado were associated with a 1E-5 probability, with a total windspeed of 132 mph and a pressure drop of 0.4 psi. The missile suite selected included a representative NRC Standard Review Plan penetrating missile and a representative crushing missile. All necessary modifications to the B Diesel Generator room, Control Building, Intermediate Building, and Standby Auxiliary Feedwater piping were completed [25].

More recently, related to high winds and especially tornado missiles, Ginna performed an extensive review of the current licensing basis and potential exposure of structures, systems, and components (SSCs) to tornado-generated missiles in response to NRC Regulatory Issue Summary (RIS) 2015-06. All vulnerabilities identified were resolved to ensure that the required SSCs remain protected, providing an acceptable resolution to the Systematic Evaluation Program (SEP) deviations identified during the review [25]. No other changes to the plant or its environs have occurred that are expected to challenge the conclusion of the IPEEE regarding high wind events. Additionally, the risk from high winds is likely to be reduced by the recent additions of FLEX equipment and procedures to the site.

b)

The evaluation of the impact of the external flooding hazard at the site was updated as a result of the NRCs post-Fukushima 50.54(f) Request for Information. The stations flood hazard reevaluation report (FHRR) was submitted to the NRC for review on March 11, 2015 [20]. The results indicated that all flood causing mechanisms, except Local Intense Precipitation (LIP) and combined effects River Flood which produces a probable maximum flood (PMF), were bounded by the current licensing basis (CLB) and did not pose a challenge to the plant.

The reevaluated LIP mechanism was found to produce various water surface elevations (WSEs) at different locations throughout the site. At the Auxiliary Building, the peak water surface elevation is 270.9 ft and the finished floor elevation is 271.0 ft. Peak LIP WSEs at the battery and diesel generator rooms are 255.8 ft with the buildings having a finished

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-26 G1-LAR-006, Rev. 1 floor elevation of 253.5 ft. Both structures have normally closed watertight doors and seals that provide 4.5 ft protection against flood water intrusion. Therefore, the available physical margin (APM) against flooding is 2.2 ft and no impacts are expected to the Auxiliary Building, Battery or Diesel Generator rooms. The peak WSE at the screen house is 255.8 ft and water intrusion is expected to affect the Service Water system, however, SW is not credited for providing cooling water during an external flood. An alternate cooling water tank is available at elevation 271.0 ft.

The PMF resulting from the combined effects river flood would inundate the site and the CLB requires temporary barriers to be installed by site personnel prior to the arrival of flood waters. As outlined in the Ginna Focused Evaluation (FE) [21], the evaluation concluded the site has an adequate site response and APM to mitigate the effects from the PMF.

To better characterize the frequency of exceedance for the combined effects river flood risk-significant flood events, a flood-frequency study was completed on August 5, 2019

[22]. The report analyzed flooding events up to an exceedance frequency of 1E-6/yr and provided inundation mapping to show the impact to the site from a flood with an exceedance frequency of 1E-6/yr. The results show that a combined effects river flood with this exceedance frequency would not produce a WSE greater than the elevation of the stream banks on the south and east sides of the plant.

2.7 RAI 7 - PLANT SPECIFIC RISK CALCULATIONS REQUEST:

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

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-27 G1-LAR-006, Rev. 1 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.

RESPONSE

a)

Table 5-15, for internal events ICCDP/ICLERP, has one case of results with a 0.00E+00 result, for case CONT-PRESS. This 0.00E+00 result is acceptable because a failed containment pressure High-High signal for containment spray initiation has very little impact on CDF or LERF. For CDF, the success of containment spray as a method for decay heat removal during high pressure recirculation is modeled, so a potential impact on CDF does exist. However, because the function can also be provided by containment fan coolers, the impact of a single signal failure is very low and falls below the 1E-12 truncation level for CDF calculations. For LERF, the same impact is seen

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-28 G1-LAR-006, Rev. 1 because containment spray is not credited to prevent a large early release for the large dry containment at Ginna. Again, the impacts fall below the 5E-13 truncation level for LERF calculations. Calculations of a specific gate (CS300Y, FAILURE TO PROVIDE FLOW FROM CONTAINMENT SPRAY DURING INJECTION) show a change in failure probability from 1.73E-2 to 1.76E-2, but this small difference does not impact the overall reported CDF or LERF due to the redundancy provided by the containment fan coolers.

Table 5-17, for Fire PRA ICCDP/ICLERP, several other cases in addition to CONT-PRESS show 0.00E+00 results (OVR-TEMP, SG-WTR-LVL-A, SG-WTR-LVL-B, RCS-FLOW-A, RCS-FLOW-B). The same explanation applies to the Fire PRA results, that any changes due to these failed components falls below the truncation level of the calculations. The same flag files are used in both the internal events PRA calculations and the Fire PRA calculations, and the events modified are confirmed in the Fire PRA model, so the changes indicated in the internal events PRA support that the changes are captured in the logic but are below the truncation levels for those specific Fire PRA cases.

b)

For the delta CDF and delta LERF calculations in Tables 5-14 and 5-16, the process for the calculations is described in Section 5.1.4 of the LAR:

New unavailability values were applied to the ESFAS/RTS instrumentation based on the change in Allowed Outage Time. The change in unavailability is assumed proportional to the change in Allowed Outage Time. These new values can be seen in Table 5-14.

The model changes for the delta CDF and delta LERF calculations are discussed in Section 5.6.4.2 of the LAR, which shows how unavailability events for each RTS/ESFAS channel are increased, including the events for CONT-PRESS and all other cases that return 0.00E+00 results in Tables 5-15 and 5-17. The flag file for the delta CDF and delta LERF calculations includes entries to increase the unavailability of the basic events for all cases. For example, for CONT-PRESS, events TS3.3.2.J(2C1)X and

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-29 G1-LAR-006, Rev. 1 TS3.3.2.J(2C2)X are increased from 0 (not included in the base PRA model) to 1.37E-3.

All other basic events for other changes are captured similarly.

In addition to the direct impacts of RTS/ESFAS components on automatic actuation signals, it is recognized that some such components may also have indirect impacts on other aspects of the PRA model. A review of the RTS/ESFAS components identified that a failure of the SG-WTR-LVL signals could impact the available level of water in the steam generators at the time of the plant trip. A significantly lower steam generator water level than expected could impact timing aspects of the Human Reliability Analysis if nominal water levels are assumed for some actions.

However, in internal events models, failures such as these are typically considered negligible due to various plant features that minimize the impact of the loss of a single signal. During an actual loss of feedwater event where the steam generator water level signals failed to provide signals, other physical parameters of the plant (e.g., RCS temperature and pressure) would also be impacted due to the loss of heat transfer, so the likelihood of multiple such signals all failing to provide the plant trip is very low. Manual trip would also be likely as other plant parameters deviate from their nominal values.

Effects of multiple spurious operations on fire initiating events have been reviewed as part of the development of the Fire PRA. Therefore, similar to the discussion above, such non-modeled impacts of RTS/ESFAS components have a negligible impact on CRMP calculations.

c)

The 0.00E+00 delta LERF presented in Table 5-16 for the fire PRA is due to the same issue described in part (a) above - the changes fall below the truncation levels.

The same flag files are used in both the CDF and LERF calculations, and the events modified are confirmed in the Fire PRA model. The changes indicated in the CDF results support that the changes are captured in the logic but are below the truncation levels for LERF.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-30 G1-LAR-006, Rev. 1 2.8 RAI 8 - TIER 3 EVALUATIONS REQUEST:

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.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 2-31 G1-LAR-006, Rev. 1 If the CRMP model does not model relevant signals and components, please describe and justify how the CRMP evaluation is performed.

RESPONSE

a)

The fire risk in support of 10CFR50.65 (a)(4) is controlled through OP-AA-201-012-1001, OPERATIONS ON-LINE FIRE RISK MANAGEMENT [23]; it is not evaluated quantitatively in PARAGON. When a fire component is added to the schedule, PARAGON alerts operators and they are directed to the procedure steps to manage risk in accordance with Exelons implementation of the qualitative fire in (a)(4) guidance in NUMARC 93-01 [24]. Compensatory measures are pre-established for any risk significant equipment removed for maintenance. Every time the fire model is updated or an issue is discovered, the list of risk significant equipment is validated or updated.

The valid actuation of the reactor trip and ESFAS signals are not a risk significant function in the fire model. For a fire to be risk significant, that same fire typically disables automatic actuation systems; this forces operations to manually trip the reactor and locally align equipment. In these scenarios, reactor trip and ESFAS signals are not credited.

b)

For the purposes of (a)(4), only pre-identified risk significant RTS and ESFAS systems are modeled in the online risk monitoring software, PARAGON. Modeled components include AMSAC (Anticipated Transient Without Scram Mitigation System Actuation Circuitry), reactor trip breakers, power supplies that support RTS/ESFAS functions, master relays, and undervoltage relays. Note that this detail exists only for the automatic response that are credited in the PRA model. The PRA model does not credit signals where it cannot be ensured that a signal exists for specific accident sequences.

Some, but not all, of the lower-level components, including those utilized in the flag files discussed in Section 5.6.4.2.2 of the LAR, are explicitly modeled in the CRMP software.

For any that are not explicitly included in PARAGON, they are assumed to have a negligible impact on CRMP calculations based on their pre-determined low risk significance.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 3-1 G1-LAR-006, Rev. 1

3.0 REFERENCES

1. Sreenivas, V, email to Thomas R Loomis, 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), 5 August 2020.

2. Exelon Generation, License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times, 25 March 2020.

3. Westinghouse Electric Company, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, WCAP-14333 Rev. 1, October 1998.

4. Westinghouse Electric Company, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, WCAP-15376 Rev. 1, March 2003.

5. TSTF-411 Surveillance Test Interval Extensions for Components of the Reactor Protection System, WOG-151, Rev 0.

6. TSTF-418 RPS and ESFAS Test Times and Completion Times (WCAP-14333),

WOG-152, Rev. 0.

7. G1-LAR-005, Ginna PRA Analysis for ESFAS/RTS AOT Extension, February 2020.

8. LTR-RAM-II-09-049, RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the R.E. Ginna Station Probabilistic Risk Assessment, Westinghouse Electric LLC, August 2009.

9. 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, American Society of Mechanical Engineers, New York, NY, February 2009.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 3-2 G1-LAR-006, Rev. 1

10. 032299-RPT-002, Revision 0, Ginna Nuclear Plant PRA Finding-Level Fact and Observation Technical Review, Jensen Hughes, August 2017.

11. G1-MISC-023, Revision 0, Ginna PRA Finding Level Fact and Observation Independent Assessment, Jensen Hughes, March 2020.

12. Close Out of Facts and Observations (F&Os). Final Version of Appendix X for NEI 05-04, NEI 07-12 and NEI 12-06 Nuclear Energy Institute, February 21, 2017.

13. NEI 05-04, Rev. 3, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Nuclear Energy Institute, November 2009.

14. ML13093A064, "R.E. Ginna Nuclear Power Plant, Docket No. 50-244, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)," Constellation Energy, March 2013.

15. ML15271A101, R.E. Ginna Nuclear Plant - Issuance of Amendment Regarding Transition to Risk Informed, Performance-Based Fire Protection Program in Accordance with Title 10 of the Code of Federal Regulations Section 50.48© (CAC No. MF1393), Nuclear Regulatory Commission, November 2015.

16. G1-PRA-012, Revision 4, Internal Flooding Analysis Notebook, February 2020.

17. US NRC, Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard in Response to March 12, 2012 50.54(f) Request for Information, 18 February 2016, ML15364A544.

18. CENG, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, 31 March 2014, ML14099A196.

19. US NRC, Safety Evaluation Report related to the full-term operating license for R.E. Ginna Nuclear Power Plant, NUREG-0944, October 1983.

G1-LAR-006 ESFAS/RTS AOT Extension - RAI Responses 3-3 G1-LAR-006, Rev. 1

20. Constellation Energy Nuclear Group, LLC Letter to USNRC, Responses to March 12, 2012 Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 2, Flooding Hazard Reevaluation Report, 11 March 11, RS-15-069.

21. Exelon Generation Company LLC, Letter to USNRC, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal, 10 March 2017, ML17069A004.

22. Attera Solutions, Flood-Frequency Analysis for Localized and Stream Flooding, R.E. Ginna Nuclear Power Plant, 20 May 2019.

23. OP-AA-201-012-1001, Operations On-Line Fire Risk Management, Revision 4.

24. Nuclear Energy Institute, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, NUMARC-93-01, Revision 4, December 2010.

25. Design Analysis DA-CE-17-001, Tornado Missile Protection Structural Barriers, Revision 000, May 2018.

1 Thornsbury, Eric From:

Falvo, Samuel Sent:

Thursday, September 3, 2020 4:29 PM To:

Thornsbury, Eric

Subject:

G1-LAE-006 Attachments:

G1-LAR-006 RAI Responses Rev 1 markup.docx I have reviewed this document and authorize my Risk Management signature.

SAM FALVO Lead Engineer l Risk Informed Engineering 1500 McConnor Parkway, Suite 500 Schaumburg, IL 60173 O: +1 630-491-8123 sfalvo@jensenhughes.com jensenhughes.com