L-2023-118, Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML23255A041
Person / Time
Site:	Saint Lucie
Issue date:	09/11/2023
From:	Strand D Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-2023-118
Download: ML23255A041 (1)

Text

September 11, 2023 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk

SUBJECT:

DOCKET NO:

St. Lucie Units 1 and 2 50-335 and 50-389 L-2023-118 10 CFR 50.90 10 CFR 50.69 Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors"

References:

1. Florida Power & Light Company letter L-2022-175, Application to Adopt 1 O CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," December 2, 2022 (ML22336A071)

2. NRC Message from Michael Mahoney, Project Manager for St. Lucie Units 1 and 2, "Request for Additional Information - St. Lucie Plant, Units 1 and 2 - Adopt 10 CFR 50.69 (L-2022-LLA-0182)," August 14, 2023 (ML23226A076)

In Reference 1, Florida Power & Light Company (FPL) submitted a license amendment request to modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR),

Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," to the licenses for St. Lucie Unit 1 and Unit 2.

Reference 2 transmitted a request for additional information that the NRC staff has determined is necessary to complete its review.

The enclosure to this letter provides the requested information.

This letter does not alter the conclusions in Reference 1 that the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change.

This letter contains no new or revised regulatory commitments.

St. Lucie Nuclear Plant Docket Nos. 50-335 and 50-389 L-2023-118 Page 2 of 2 If you should have any questions regarding this submittal, please contact Kenneth Mack, Licensing Manager, at (561) 904-3635.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 11 th day of September 2023.

General Manager Regulatory Affairs Florida Power & Light Company

Enclosure:

Evaluation of the Proposed Change cc:

USNRC Regional Administrator, Region II USNRC Project Manager, St. Lucie Nuclear Plant, Units 1 and 2 USNRC Senior Resident Inspector, St. Lucie Nuclear Plant, Units 1 and 2 Mr. John Williamson, Florida Department of Health

ENCLOSURE Response to Request for Additional Information (21 pages follow)

REQUEST FOR ADDITIONAL INFORMATION (RAI)

LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FLORIDA POWER & LIGHT COMPANY ST. LUCIE, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 389 EPID L-2022-LLA-0182 RAI-01 (APLA) - Credit for FLEX Equipment and Actions NRC memorandum dated May 6, 20221, provides the NRCs staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.2002. The staff considers the May 6, 2022, memorandum to be applicable to any other portable equipment credited in PRA models.

With regards to equipment failure probability, in the memorandum dated May 6, 2022, the NRC staff states in Conclusion 4:

Licensees that choose not to use the generic failure probabilities in PWROG-10842 to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRA used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval.

With regards to the uncertainty related to equipment failure probabilities, in the updated NRC memorandum, the NRC staff states in Conclusion 8:

PWROG-18043, Revision 1, notes that there was insufficient data to quantify the failure to load probabilities for portable diesel generators due to lack of detailed data. To account for the uncertainty in the testing activitieslicensees should ensure their preventive maintenance strategies include such testing and that the data reported provides this information. licensees should continue to assess the uncertainty in equipment failure rates and address or disposition it.

With regards to HRA, in the memorandum dated May 6, 2022, the NRC staff states, in part, in Conclusion 11:

1U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments, dated May 6, 2022 (ADAMS Accession No. ML22014A084).

2U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, RG 1.200, Revision 3, December 2020 (ADAMS Accession No. ML20238B871).

EPRI 3002013018 provides updated detailed industry guidance for estimating the human error probabilities (HEPs) of the actions needed to implement mitigating strategies using portable equipment. EPRI 3002013018 provides guidance that is acceptable to the NRC, with the clarifications below...

With regards to PRA Upgrade, the staff states in the update memorandum in Conclusion 2:

Therefore, Conclusion 2 remains unchanged [that] for any new risk-informed application that has incorporated mitigating strategiesthe licensee should either perform a focused-scope peer review of the PRA model or demonstrate [that it does not meet the three criteria of an PRA Upgrade].

The NRC staff understands the St. Lucie PRA models does not incorporate FLEX equipment and mitigation strategies, but includes other portable equipment in the PRA models used for this application.

a) Clarify if the St. Lucie PRA models credit any portable equipment during the categorization process. If portable equipment is credited, then respond to the following parts.

b) Describe the methodology used to assess the failure probabilities of any modeled portable equipment credited in the licensee's PRA model. The discussion should include a justification of the rationale for parameter values, and how the uncertainties associated with the parameter values are considered in the categorization process in accordance with ASME/ANS RA-Sa-20093, as endorsed by RG 1.200 (e.g., supporting requirements for HLR-DA-D).

c) A discussion detailing the methodology used to assess operator actions related to portable equipment and the licensee personnel that perform these actions.

The discussion should include:

i. A summary of how the licensee evaluated the impact of the NRC clarification with regards in using the EPRI 3002013018 FLEX HRA methodology.

ii. Provide updated portable equipment HRA results, if required, to address the NRC clarifications.

iii. Provide justification that the use of the EPRI FLEX HRA methodology does not meet the definition of an PRA Upgrade as defined by RG 1.200.

-OR-Propose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the EPRI FLEX HRA method for the St. Lucie

3American Society of Mechanical Engineers (ASME) and American Nuclear Society (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", February 2009, New York, NY (Copyright).

PRA models. Include in the mechanism to close out all F&Os that result from the FSPR prior to implementing the categorization process.

d) Provide an assessment, such as a sensitivity study, of the impact on risk associated with the uncertainty in portable equipment and operator action failure rates credited in St. Lucies PRA models. This assessment should include, if required, any modifications to portable equipment modeling based on the issues raised in this question. Include in this discussion the impact of SSC risk importance associated with the uncertainty in portable equipment and operator action failure rates on the categorization process (e.g., an SSC goes from low safety significance in the base case to high safety significance in the sensitivity case).

RESPONSE

a) The PRA credits a diesel driven portable fire pump (B.5.b), hence referred to as portable pump, in its PRA. Any FLEX equipment credited in the future will apply NRC accepted data sources and HRA methodologies as a prerequisite to implementing the 10 CFR 50.69 categorization processes.

b) The PRA currently utilizes industry generic failure probabilities for engine-driven pumps from NUREG/CR-6928, INL/EXT-21-65055, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants: 2020 Update, which are plant specific Bayesian updated for modeling failure probabilities of this portable pump. As a prerequisite to implementing the 10 CFR 50.69 categorization processes, FPL will update these failure probabilities modeled in its PRA and apply NRC accepted data sources.

c) An operator action to utilize the portable pump for make-up to the CST is currently credited in the PRA, with a screening 0.1 human error probability. As a prerequisite to implementing the 10 CFR 50.69 categorization processes, FPL will update the operator action for this portable pump by utilizing NRC accepted HRA methodologies and perform a FSPR of this update followed by any necessary F&O closure items by an Independent Assessment Team IAT per NEI 05-04/07-12/12-06 Appendix X: Close-Out of Facts and Observations (F&Os).

d) As a sensitivity to inform risk ranking related to 10 CFR 50.69, Table 1, provides CDF and LERF Fussel Vesley (FV) and Risk Achievement Worth (RAW) ranking of equipment-related basic events for the Unit 1 base model compared to a sensitivity model with the B.5.b HEP set to TRUE which changed risk significance (i.e., High to Low or Low to High).

As can be seen in Table 1, the B.5.b pump has a modest impact on risk ranking and the base model is judged to provide a reasonable and conservative treatment of actual plant capabilities. Altogether 20 components changed risk significance (FV>0.005=HIGH or RAW>2=HIGH) with only 4 components going from Low to High shown in top 4 rows and the remainder going from High to Low with disposition for these being shown in the rightmost column. That column explains the impact of the change in risk significance of the event. The events showing slight increase in RAW are already important by other related component events associated with that component (e.g., Test and Maintenance

term for HPSI pump A went to HIGH but its failure to run event was already HIGH in the base case and for the EDG CCF, the individual EDG failure to run is already HIGH). For those going from High to Low the change is small and would not warrant a recommendation to change the components risk categorization.

Table1 Event Description Fus Ves (CDF Base)

AchW (CDF Base)

Fus Ves (LERF Base)

AchW (LERF Base)

Sig.Base Fus Ves (CDF Sens)

Ach W

(CDF Sens)

Fus Ves (LERF Sens)

AchW (LERF Sens)

Sig.Sens Disposition GTM1PUMPA HPSIPUMPAINTEST ORMAINTENANCE 0.004 1.85 0.001 1.22 LOW 0.005 2.02 0.002 1.52 HIGH GMPF1PAINJ"HPSI PUMP1AFAILSTORUN DURINGINJECTION" already"HIGH"inBase EMMDGRABCD CCF(4/4)OFEDG 1A,1B,2A,AND2BTO RUN 0.000 1.89 0.000 1.65 LOW 0.000 2.02 0.000 1.78 HIGH EDGs(e.g.,

EDGR11BEDG)already "HIGH"inBase;CCF RAW<20.

OMM1CPORVO CCFOF2/2PORVSTO OPEN 0.000 1.33 0.000 1.59 LOW 0.000 2.95 0.000 3.14 HIGH PORVFTO(e.g.,

ORZN1V1402, ORZN1V1404)already "HIGH"inBase;CCF RAW<20.

EMM_INVA_CCCF COMMONCAUSE FAILUREOF INSTRUMENTBUSINV MA&MC 0.000 1.93 0.000 1.00 LOW 0.000 2.00 0.000 1.00 HIGH SimilarInstrumentBus CCFcombinationcases already"HIGH"and onlyasingle,verysmall CDFRAWdelta.

AMVN1083 MOTOROPERATED VALVEMV083FAILS TOOPEN 0.004 2.01 0.005 2.18 HIGH 0.003 1.79 0.003 1.87 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" ATPF1AFW1C4 TURBINEDRIVENAFW PUMP1CFAILSTO RUN4HR 0.003 1.99 0.004 2.09 HIGH 0.003 1.78 0.003 1.80 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW"

ATPA1AFW1C TURBINEDRIVENAFW PUMP1CFAILSTO START 0.002 2.00 0.002 2.15 HIGH 0.001 1.79 0.001 1.84 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GMPF1PBREC HPSIPUMP1BFAILS TORUNDURING RECIRCULATION 0.002 2.24 0.000 1.20 HIGH 0.001 1.98 0.000 1.15 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" AXVK109120 MANUALVALVE V09120TRANSFERS CLOSED 0.001 1.77 0.001 2.11 HIGH 0.000 1.60 0.001 1.77 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" CAVC1HCV1410 AIROPERATEDVALVE HCV1410FAILSTO CLOSE 0.001 2.00 0.000 1.16 HIGH 0.000 1.79 0.000 1.11 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" CAVC1HCV148B AIROPERATEDVALVE HCV148BFAILSTO CLOSE 0.001 2.00 0.000 1.16 HIGH 0.000 1.79 0.000 1.11 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" NLCD1AM606 LOGICCIRCUITAM606 FAILSTOGENERATE SIGNAL 0.000 2.00 0.000 1.15 HIGH 0.000 1.79 0.000 1.11 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GCVK13410R CHECKVALVEV3410 TRANSFERSCLOSED DURING RECIRCULATION 0.000 2.22 0.000 1.17 HIGH 0.000 1.97 0.000 1.12 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GCVK13414R CHECKVALVEV3414 TRANSFERSCLOSED DURING RECIRCULATION 0.000 2.22 0.000 1.17 HIGH 0.000 1.97 0.000 1.12 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW"

GCVN107172 CHECKVALVEV07172 FAILSTOOPEN 0.000 2.21 0.000 1.13 HIGH 0.000 1.98 0.000 1.12 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GTM1MV072B MV072BTESTAND MAINTENANCE 0.000 2.18 0.000 1.12 HIGH 0.000 1.96 0.000 1.10 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" CMVR1141FO MOTOROPERATED VALVEMV141 TRANSFERSOPEN 0.000 2.19 0.000 1.00 HIGH 0.000 1.94 0.000 1.00 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GMVK13654R MOVV3654 TRANSFERSCLOSED DURING RECIRCULATION 0.000 2.12 0.000 1.00 HIGH 0.000 1.89 0.000 1.00 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GMVK1MV072B MOVMV072B TRANSFERSCLOSED 0.000 2.13 0.000 1.00 HIGH 0.000 1.91 0.000 1.00 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW" GXVK13411R MANUALVALVE V3411TRANSFERS CLOSEDDURING RECIRCULATION 0.000 2.17 0.000 1.06 HIGH 0.000 1.92 0.000 1.04 LOW SlightreductioninRAW (below2.0),not recommendedto reclassas"LOW"

RAI-02 (APLA) - Determination of Key Sources of Uncertainty for the 10CFR50.69 Categorization Process and Sensitivity Results Sections 50.69(c)(1)(i) and 50.69(c)(1)(ii) of 10 CFR require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience. The guidance in NEI 00-04 specifies that sensitivity studies be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the importance of components. The guidance in NEI 00-04 states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

Section 3.2.8 of the LAR Enclosure describes the process used for reviewing the PRA assumptions and sources of uncertainty. The NRC staff reviewed the uncertainty documents provided on this audits electronic portal for the internal events, internal flooding, and fire PRA and found that further clarification is necessary regarding the review of assumptions and sources of uncertainty for this application. It is unclear if additional analysis was performed and documented to determine if any source of uncertainty could adversely impact any SSC categorization. Some portal documents referred to sensitivity studies that are contained in other documents, however these sensitivity results were not provided on the portal. In light of these observations, provide the following information:

a) Provide details of how the PSL PRA sources of uncertainty were evaluated as a potential key source of uncertainty for this application. Include in this discussion any documentation of this process.

b) Provide the results of sensitivity studies that determined the impact on risk for each associated source of uncertainty. Include in this discussion justification that the sensitivity results demonstrate that the associated source of uncertainty does not adversely impact any SSC categorization.

RESPONSE

a) The detailed process of identifying, characterizing, and qualitative screening of model uncertainties that was used for the St. Lucie PRA is documented in Section 5.3 of NUREG-1855 Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, and Section 3.1.1 of EPRI TR-1016737, December 2008, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments. PSL-SNBK-UNC, Revision 1, Uncertainty Notebook for St. Lucie Units 1 and 2, documents the results of the uncertainty analysis. For Fire PRA, the Uncertainty assessment is also guided by EPRI-1011089-NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, as documented in PSL-BFJR-16-050, Revision 1, St. Lucie Nuclear Plant Uncertainty and Sensitivity Report NUREG/CR-6850 Task 15.

A number of sensitivities are directed by 10 CFR 50.69-related Fleet Procedure EN-AA-112-1000, Revision 1, RIEP Active Component Risk Significance Insights, including:

Increase all human error basic events to their 95th percentile value Decrease all human error basic events to their 5th percentile value Increase all component common cause events to their 95th percentile value Decrease all component common cause events to their 5th percentile value Set all maintenance unavailability terms to 0.0 For the LAR, the list of uncertainties compiled in the PSL Uncertainty Notebooks was reviewed to identify which of the assumptions or sources of uncertainty could significantly impact the risk calculations that support the SSC categorization process.

Based on this review, which is documented in Attachment 7 of PSL-BFJR-17-056, Revision 1, PSL PRA Input to the 50.69 LAR, those assumptions or sources of uncertainty that were determined to have the potential to significantly impact categorization of SSCs were considered key for the LAR.

of the LAR lists three key sources of modeling uncertainty:

i.

Ignition Source Counting - as noted in Attachment 6, this uncertainty is addressed by using an industry accepted consensus modeling approach. No sensitivity assessments are required.

ii.

Generic Fire Modeling - as noted in Attachment 6, this uncertainty is addressed by using an industry accepted consensus modeling approach. No sensitivity assessments are required.

iii.

Human-Induced errors in support system initiating event models - as noted in, this uncertainty has the potential to affect categorization of SSCs and the model will be updated to remove the key uncertainty or a sensitivity assessment will be provided as a prerequisite to implementing the 10CFR 50.69 categorization processes.

As discussed during the NRC audit, for internal events and flooding hazards, the Uncertainty Notebook for St. Lucie Units 1 and 2, PSL-SNBK-UNC, Revision 1, documents the results of the PSL PRA uncertainty analysis in Table 48 Issue Characterization for St. Lucie-Specific Sources of Model Uncertainty and Table 49 Issue Characterization for EPRI Generic Sources of Model Uncertainty. The PSL PRA Input to the 10 CFR50.69 LAR, PSL-BFJR-17-056 Revision 1, reviews each source of uncertainty identified in Tables 48 and 49. Attachment 7 to PSL-BFJR-17-056 Revision 1 provides an assessment of each item documented in the PSL-SNBK-UNC Revision 1 that is identified in Table 48 and Table 49 as shown in the rightmost column titled 50.69 application assessment (items highlighted in aqua are passed up to attachment 6). Four total were colored aqua. One of the items, which is identified as item iii above, was considered a key source of uncertainty which potentially affects the risk categorization of SSCs and is therefore added to attachment 6 of the LAR. The other three identified itemswere erroneously left as aqua colored and were not considered a key source of uncertainty for this application and not included in the LAR.

b) As discussed during the NRC audit, no additional sensitivity studies have been performed to date to evaluate the impact of assumptions and sources of uncertainty of the SSC categorization for PSL. The key uncertainty involving Human-Induced errors in support system initiating event models has the potential to affect categorization of SSCs

and the model will be updated to remove the key uncertainty as a prerequisite to implementing the 10 CFR 50.69 categorization processes. As mentioned in part a) above, standard sensitivity studies, to be performed as a minimum, are included in NextEra procedure EN-AA-112-1000, Revision 1, which is intended to determine risk significance of active structures, systems, and components (SSCs) in accordance with 10 CFR 50.69.

RAI-03 (APLA) - Open F&O CS-B1-01 Concerning Circuit Coordination Regulatory Guide (RG) 1.200, Revision 3 An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities, (ADAMS Accession No. ML20238B871), provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process utilizing the ASME/ANS PRA standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. The primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os. A process to close-out Finding-level F&Os is documented in NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (ADAMS Accession No. ML19231A182) that is endorsed by RG 1.200, Revision 3.

LAR Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items presents a Finding-level F&O (i.e., CS-B1-01) which states:

No evaluation was performed to verify that the new components and cables associated with the Fire PRA is bounded by existing overcurrent coordination analysis.

The St. Lucie disposition for Finding CS-B1-01 states:

This finding has been resolved, but independent review to certify closure has not yet been completed. This has no impact on 10 CFR 50.69 implementation.

LAR Section 3.3 explains that in April 2019, the Pressurized Water Reactor Owners Group performed an independent assessment and certified that all F&Os generated for PSL were closed except for fire-related items in Attachment 3. Since April 2019, it appears there have been efforts to resolve Finding CS-B1-01.

In the file identified as AR for None-Safety Related MCC 1A3 & 1B3 Breaker Coordination.pdf, it is stated that during review of coordination calculation, PSL-1FSE-09-001, it was discovered there were feeder breakers for loads downstream of non-safety related motor control center (MCCs) 1A3 and 1B3 modeled in the fire PRA that were not fully coordinated. In another file identified as AR 2318093 Enhance Non-Safety MCC BRKS Coord for the Fire PRA Modeling, it is stated for MCC 1A3 and 1B3 that enough margin exists to shift their current time current characteristics (TCC) to the right without challenging coordination with the upstream load center breakers. This statement appears to indicate that some adjustment is needed to the TCC of these breakers to resolve the coordination issue. It is not completely clear to NRC whether the adjustment has been made to resolve the coordination issue associated with MCC 1A3 and 1B3, which are credited in the PRA, has been resolved.

NRC staff notes that circuit breakers and fuses should be adequately coordinated with the upstream load center breaker over the rated range of the circuit, to prevent the adverse effects of a fault on the rest of the circuits powered from a common source. If circuits modelled in the

fire PRA cannot be found to be coordinated, then the coordination issues should be physically resolved, or the negative effects of the coordination issue should be modelled in the fire PRA.

It appears that an evaluation has been performed and was documented in a cited report (i.e.,

PSL-1FSE-09-001) to evaluate whether new components and cables associated with the Fire PRA are bounded by existing overcurrent coordination analysis as requested in Finding CS-B1-

01. In the file identified as AR 2318092 NRC TFP1 SR MCCS Potential Lack of Coordination.pdf, it is stated that three generic sources of potential weaknesses in circuit coordination on Unit 1 Safety-Related MCCs were identified but were shown not to be a concern to safe shutdown or impact PRA risk. Accordingly, it appears that Safety-Related MCCs are not a concern but certain non-Safety-Related MCCs are a concern. However, the full scope content of PSL-1FSE-09-001 is not known.

Given the observations above, address the following:

a) Confirm that Finding CS-B1-01 has been closed using an NRC approved F&O closure process,

- OR -

b) Commit to a licensee condition (e.g., an implementation item) that ensures Finding CS-B1-01 will be closed using an NRC approved F&O closure process prior to implementation of the 10 CFR 50.69 risk categorization program.

- OR -

c) Describe the evaluation that was performed to verify that the new (non-safe shutdown) components and cables modeled in the fire PRA are bounded by existing overcurrent coordination analysis. Also, describe actions (if any) performed after the evaluation to ensure circuit coordination. Include in this description:

i.

Discussion of how the evaluation assures that new (non-safe shutdown) components and cables modeled in the Fire PRA are bounded by existing overcurrent coordination analysis.

ii.

Discussion of the results of the evaluation discussed above.

iii.

If certain circuits were found to uncoordinated by the evaluation, then describe the efforts undertaken that resolve the coordination issue(s), or demonstrate (e.g., through a sensitivity study) that the impact of the coordination issues do not have a consequential impact on 10 CFR 50.69 risk categorization.

RESPONSE

The approach to be taken for resolution of F&O CS-B1-01 is similar to the approach proposed for resolution of RAI-04 (APLA) (related to F&O CS-A3-01).

Since the associated F&O CS findings have not been formally closed using an NRC approved F&O closure process, F&O CS-B1-01 along with CS-A3-01 will be closed using an NRC approved F&O Closure process as a prerequisite to implementing the 10 CFR 50.69 categorization processes.

RAI-04 (APLA) - Open F&O CS-A3-01 Concerning MSO of Fire PRA Components Regulatory Guide (RG) 1.200, Revision 3 An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities, (ADAMS Accession No. ML20238B871), provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process utilizing the ASME/ANS PRA standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. The primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os. A process to close-out Finding-level F&Os is documented in NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (ADAMS Accession No. ML19231A182) that is endorsed by RG 1.200, Revision 3.

LAR Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items presents a Finding-level F&O (i.e., CS-A3-01) which states:

4kV power and 125VDC control cables required to support the operation of the Containment Spray Pump were not identified. Fire PRA Plant Response model and other Fire PRA support tasks are adversely affected. Perform a comparison of the components identified on the MSO (multiple spurious operation) list against the Fire PRA components for which new cable selection was performed (i.e., components not previously identified on the Appendix R safe shutdown equipment list. Verify that the cable selection for the common components supports all credited operations. Fire PRA Plant Response model and other Fire PRA support tasks are adversely affected.

The St. Lucie disposition for Finding CS-A3-01 states:

This finding has been resolved, but independent review to certify closure has not yet been completed. This has no impact on 10 CFR 50.69 implementation.

Again, LAR Section 3.3 explains that in April 2019, the Pressurized Water Reactor Owners Group performed an independent assessment and certified that all F&Os generated for PSL were closed except for fire-related items in Attachment 3. Since April 2019, it appears there have been efforts to resolve the Finding CS-A3-01.

In license report PSL-BFJR-16-039 (non-public) St Lucie NFPA 805 - Task 2 Component and Cable Selection, dated August 24, 2020, it is explained that to address Multiple Spurious Operations (MSO) for the fire PRA a review using an expert panel process is performed to identify and characterize potential MSO combinations that lead to new accident sequences. In license report FPL-SL120-PR-001 (non-public), Update Review for St. Units 1 and 2 Cable-to-Fire-to-Compartment Relationships, an MSO designator is a assigned a large fraction of the plant components. However, NRC staff could not conclude after reviewing these two reports that the circuits of concern cited in the Finding had been identified (i.e., 4kV power and 125VDC control cables required to support the operation of the Containment Spray Pump). Moreover, NRC staff could not confirm whether components in the MSO list were compared to the fire PRA component list to identify additional circuits that may need to be selected or ensure that cable selection for common components supports all credited operations. Given the observations above, address the following:

a) Confirm that Finding CS-A3-01 has been closed using an NRC approved F&O closure process,

-OR -

b) Commit to a licensee condition (e.g., an implementation item) that ensures Finding CS-A3-01 will be closed using an NRC approved F&O closure process prior to implementation of the 10 CFR 50.69 risk categorization program.

OR c) Describe the evaluation that was performed to 1) compare the components identified on the MSO list to the fire PRA component list to identify additional circuits that may need to be selected, 2) ensure that cable selection for common components supports all credited operations, and 3) identify the 4kV power and 125VDC control cables required to the operation of the Containment Spray Pump were identified for the fire PRA.

RESPONSE

A review of the current St. Lucie Unit 1 and Unit 2 Fire PRA model confirms that the CS 1A, 1B, 2A and 2B power and control cables are included in the analysis and are mapped to basic events associated with failure of the CS pumps.

Since this F&O finding has not been formally closed using an NRC approved F&O closure process, Finding CS-A3-01 will be closed using an NRC approved F&O Closure process as a prerequisite to implementing the 10 CFR 50.69 categorization processes.

RAI-05 (APLA) - Status of Fire F&Os Regulatory Guide (RG) 1.200, Revision 3 An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities, (ADAMS Accession No. ML20238B871), provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process utilizing the ASME/ANS PRA standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. The primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os. A process to close-out Finding-level F&Os is documented in NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (ADAMS Accession No. ML19231A182) that is endorsed by RG 1.200, Revision 3.

The NRC staff reviewed the peer review and closure review documents provided on this audits electronic portal regarding open F&Os and found that further clarification is necessary regarding the status of three fire F&Os. The licensee document PSL-FJR-17025 (non-public) regarding the January 2010 fire PRA peer review states in Section 4 and Table 4-15 that a total of thirty-six findings were determined by the review team. The NRC staff notes that the documents PSL-BFJR-18-020 and PSL-BFJR-19-005 (non-public) state that only thirty-three fire PRA findings were reviewed on both occasions. It is unclear to the staff the correct status of the F&Os closed out after the August 2018 re-review.

a) Provide clarification of the F&Os (all models) that were closed out after the August 2018 review.

b) Identify, if any, F&Os that were not assessed as closed by a closure review team.

Include in this discussion the disposition of these F&Os for this application.

RESPONSE

To clarify the status of open F&Os and to implement a closure review team assessment of F&Os not formally closed, a review of open F&Os and closure of these F&Os will be implemented. FPL will generate an F&O disposition matrix with reference to the closure review that closes each F&O and references the NRC approved F&O Closure process used to close the F&O. Closure of all F&Os will be performed as a prerequisite to implementing the 10 CFR 50.69 categorization processes.

RAI-06 (APLA) - Status of PRA Upgrades Associated with F&Os Regulatory Guide (RG) 1.200, Revision 3 An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities, (ADAMS Accession No. ML20238B871), provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process utilizing the ASME/ANS PRA standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. The primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os. A process to close-out Finding-level F&Os is documented in NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (ADAMS Accession No. ML19231A182) that is endorsed by RG 1.200, Revision 3.

The NRC staff reviewed the peer review and closure review documents provided on this audits electronic portal regarding open F&Os and found that further clarification is necessary regarding the disposition of closed F&Os that resulted in PRA Upgrades. The NRC staff notes that the August 2018 closure review (PSL-BFJR-19-005, non-public) states that the independent assessment team (IAT) determined that four fire F&Os (F-5 (ES-C1-01), F-6 (ES-CW-01), F-8 (FQ-C1-01), and F-24 (HRA-C1-01) remained open since the associated changes to the PRA model constituted a PRA Upgrade that required a focused-scope peer review (FSPR). The subsequent IAT report PSL-BFJR-19-024 (non-public) issued in April 2019 identified the F-5, F-6, F-8, and F-24 F&Os as PRA Maintenance. The 2019 IAT appears to have performed an FSPR on three different closed fire F&Os (IGN-A5-01, SF-A1-01, and FSS-H1-01). It is unclear to the NRC staff the inconsistency between the two IATS concerning the F-5, F-6, F-8, and F-24 F&Os.

a) Provide a description of the F&Os, the associated model changes for addressing the F&Os, a summary of the IAT evaluation of these F&Os from the August 2018 and the April 2019 F&O closures.

b) Provide clarification of the PRA model changes associated with the closure of the F-5, F-6, F-8, and F-24 F&Os and detailed justification why these changes do not constitute a PRA upgrade. Include in this discussion an explanation on the two different IAT assessments.

c) Propose a mechanism, if any of the four F&Os were determined to be PRA Upgrades, to perform a FSPR and close any associated F&O prior to implementation of the categorization process.

RESPONSE

a) Table below provides excerpts of the F&Os, description, model changes or resolutions by FPL, from 2018 IAT assessment (PSL-BFJR-19-005, Rev. 0, St. Lucie IE, IF, Fire Finding Closure Report - ENERCON) and 2019 IAT assessment (PSL-BFJR-19-024, rev. 0, Independent Assessment of Facts & Observations Closure of the Plant St.

Lucie Probabilistic Risk Assessment).

F&O F&O Summary of Assessment (PSL-BFJR-17-025, Rev. 0, Probabilistic Risk Assessment Group Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the St. Lucie, Units 1 and 2 Fire Probabilistic Risk Assessment)

Resolution (PSL-BFJR-19-005, Rev. 0, St. Lucie IE, IF, Fire Finding Closure Report - ENERCON, and/or PSL-BFJR-19-024, Rev. 0, Independent Assessment of Facts & Observations Closure of the Plant St. Lucie Probabilistic Risk Assessment) 2018 IAT assessment OPEN/UPGRADE (PSL-BFJR-19-005, Rev. 0, St. Lucie IE, IF, Fire Finding Closure Report -

ENERCON) 2019 IAT assessment CLOSED/MAINTENANCE (PSL-BFJR-19-024, Rev. 0, Independent Assessment of Facts

& Observations Closure of the Plant St. Lucie Probabilistic Risk Assessment)

ES-C1-01 (F-5)

SR NOT MET

1. Tables 4.2-1, 4.2-2 (to be completed for Unit 2), B-1 and B-2 provide information on instrumentation associated with PRA basic events and SSEL mapping and disposition. The HRA Evaluation Report (Report 0493060006.102, Rev. 0),

Tables A-1 through A-4 and Appendix C provide information on the instrumentation associated with important control room actions. Appendix R instrumentation is specifically identified by bold formatting. However, no information was provided that would allow the impact of a specific fire on the instrumentation set to be identified. For essential instrumentation this information is available in the Response to Fire procedures. The reduced set of instrumentation associated with a fire zone should be used to support estimation of the human failure probabilities associated with a fire scenario.

The instrumentation associated with specific operator actions is identified in the Component and Cable Selection notebook (PSL-BFJR-16-039); these cables were then evaluated using the process described in Section 3.0 of the Human Failure Evaluation Report (PSL-BFJR-16-041). Operator actions were evaluated to create a list of cues/instruments and verify sufficient safe shutdown related instrumentation was available to provide indications/cues to perform the credited operator actions. The availability of SSD instrument cues is assured via the safe shutdown analysis on a fire area basis. Appendix A documents this review and lists the operator actions and available cues. As part of the resolution, instrumentation logic was added adjacent to operator actions to fail the necessary cue or indication.

The change regarding impact of the circuit analysis on available instrumentation and the impact of limited instrumentation on the HEPs is a potentially risk significant change. Hence, this is a PRA upgrade. This F&O should remain open.

SR ES-C1 is MET (CC-I/II/III) since instrumentation that is relevant to operator actions with respect to FPRA is identified. No new methodology used. Change to documentation only.

ES-C2-01 (F-6)

SR NOT MET No information was identified in the Component and Cable Selection Report (Report 0493060006.101, Rev. 1) or the HRA Evaluation Report (Report 0493060006.102, Rev. 0) that characterized instrument availability or spurious operability for individual fires. See SR ES-C1 for additional information.

Provided clarification in HRA report, Section 3.

An operator does not take immediate action on a single instrument indicator or annunciator. The operator will confirm the signal using another instrument/indicator to verify that the signal is valid prior to taking action. Per ASME/ANS Standard RA-S-2008, SR ES-C2, Capability Category II; only one spurious instrument operation is assumed. Therefore, no adverse impact of errors of commission require evaluation given operator confirmation of instrumentation with other instruments prior to initiation of an action. The HRA update, prepared to support the responses to RAIs PRA 01.d, 01.h, 01.l and 01.o, addresses errors of omission The change regarding impact of the circuit analysis on available instrumentation and the impact of limited instrumentation on the HEPs is a potentially risk significant change. Hence, this is a PRA upgrade. This F&O should remain open.

SR CCII. PRA No new methodology used. Change to documentation only.

F&O F&O Summary of Assessment (PSL-BFJR-17-025, Rev. 0, Probabilistic Risk Assessment Group Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the St. Lucie, Units 1 and 2 Fire Probabilistic Risk Assessment)

Resolution (PSL-BFJR-19-005, Rev. 0, St. Lucie IE, IF, Fire Finding Closure Report - ENERCON, and/or PSL-BFJR-19-024, Rev. 0, Independent Assessment of Facts & Observations Closure of the Plant St. Lucie Probabilistic Risk Assessment) 2018 IAT assessment OPEN/UPGRADE (PSL-BFJR-19-005, Rev. 0, St. Lucie IE, IF, Fire Finding Closure Report -

ENERCON) 2019 IAT assessment CLOSED/MAINTENANCE (PSL-BFJR-19-024, Rev. 0, Independent Assessment of Facts

& Observations Closure of the Plant St. Lucie Probabilistic Risk Assessment) due to loss of instrument signals. This is accounted for by incorporation of instrument cues into the Fire PRA model fault tree, thereby, failing the operator action if the associated cues are failed by fire.

FQ-C1-01 (F-8)

SR MET Fire-related SSD actions are currently modeled only through the Altered Events file in FRANC, which bypasses the dependency analysis.

Incorporated multipliers applied to cutsets with multiple screening HEPs. See Section 4.1 and Appendix B of HFE Report. PSL-BFJR-16-041. The use of the Altered Events feature of FRANC for HRA is no longer done in the PRA model, the fault tree model now directly models the fire related human actions ensuring that fire related actions are assessed in the dependency analysis.

Appendix D describes a dependency analysis using a floor value of 1E-05.

Appendix E documents the methodology used to address dependencies in the fire PRA.

The extent of the change may be risk significant and considered an upgrade.

Also, the fire HRA dependency analysis did not exist at the time of the peer review, so it was not evaluated by the peer review team. This F&O should remain open.

SR Met. PRA The dependency analysis in FPRA utilized the same methodology for dependency analysis as was used in the Internal Events PRA model.

HRA-C1-01 (F-24)

SR CC-I Screening HEP quantification was used to adjust the existing internal event PRA to account for fire impacts. This included feasibility factors (cues Availability, accessibility of local action) and adjustment factors based on time available and complexity. This approach is appropriate for the stage of the FPRA. To achieve Cat 2, however, detailed analyses are required for the significant HFEs.

Human Failure Evaluation Report (PSL-BFRJ-16-041, Rev. 0)

Appendix B documents the new fire specific HFEs. HRA Calculator is used to define the new values for combination event recoveries given these revised base HEP values.

Further, the SR requires that the fire effects on the HEPs be quantified.

Because Screening values and multipliers were used at the time of the peer review, the methodology applied in the F&O resolution has not been reviewed and is an upgrade.

SR MET CC-II since detailed analyses for almost all of the HFEs. The HRA Report (PSL-BFRJ-16-041, Rev. 1)

Section 5.0 lists the new fire specific HFEs, which are determined and documented in the HRA Calculator.

All of the new actions have detailed calculations. No screening values are used.

b) As discussed during the NRC Audit and as noted in the August 2018 Closure Review (PSL-BFJR-19-005, Revision 0, St. Lucie IE, IF, Fire Finding Closure Report - ENERCON) FPL determined that four F&Os were upgrades [ES-C1-01 (F-5), ES-C2-01 (F-6), FQ-C1-01 (F-8), and HRA-C1-01 (F-24)]. The 2018 Independent Assessment Team (IAT) noted that upgrade was assigned because:

The four F&Os identified as PRA upgrades all appear to have satisfactorily addressed the technical modeling issues noted in the F&Os. However, the resolutions either resulted in new information that was not reviewed by a peer review team or the changes resulted in potentially significant changes in the calculated risk results and insights. As the technical issues were already corrected, these should have no impact on calculated results.

These F&Os werent directly subjected to Focused Scope Peer Review (FSPR) but rather the related Supporting Requirements (SRs) were re-assessed by the Independent Assessment team (IAT). Specifically, Section 4 of the 2019 Closure Review (PSL-BFJR-19-024 Revision 0, Independent Assessment of Facts & Observations Closure of the Plant St. Lucie Probabilistic Risk Assessment) notes:

During the F&O closure review, 4 unique F&Os were judged to be closed with a PRA upgrade, which required a focused scope peer review. This triggered a focused scope peer review of the Supporting Requirements associated with the upgrade.

Table 3-7 of the 2019 Closure/FSPR Review shows the FSPR summary for each SR in the scope of this review; including the 4 SRs related to the F-5, F-6, F-8, and F-24 F&Os. For completeness, the IAT closed the F&Os noting Maintenance to convey that no additional FSPR is required. Note that FQ-C1 is N/A for the FSPR because that SR was previously classified as Met. This IAT similarly closed the related F&O (FQ-C1-01).

c) Per the NRC Audit discussion, F-24(FQ-C1-01) will have an FSPR of this update followed by any necessary F&O closure items by an IAT per NEI 05-04/07-12/12-06 Appendix X: Close-Out of Facts and Observations (F&Os), as aprerequisite to implementing the 10CFR 50.69 categorization processes.

RAI-07 (APLC) - Seismic Tier 1 GMRS vs SSE Criteria Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation are adequate for the categorization of SSCs.

Section 3.2.3 of the LAR Enclosure references the EPRI 3002017583 Tier 1 criteria for the Ground Motion Response Spectrum (GMRS) of below or approximately equal to the Safe Shutdown Earthquake (SSE) between 1.0 and 10.0 Hertz. The LAR continues by stating that the St. Lucie response to the NRC 50.54(f) letter, regarding post-Fukushima recommendations, concluded that the plant SSE exceeded the GMRS in the specified frequency range. However, the staff notes that no curves showing the SSE and GMRS and their comparison are provided in the LAR.

Provide an SSE/GRMS hazard curve comparison demonstrating that the Tier 1 criteria are met for the appropriate frequency band.

RESPONSE

A comparison of the SSE vs GMRS curve is shown below. The curve was taken from Figure 3.4-1 of the NRC Staff Assessment of the reevaluated seismic hazard for St. Lucie (ML15352A053).

RAI-08 (APLC) - External Hazards Screening Section 2.3.1, Item 7, of NEI 06-09-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The NRC staffs SE for NEI 06-09 states that

[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.

In Attachment 4 (External Hazards Screening) of the LAR Enclosure, the licensee screens the External Flood hazard as C1, Event damage is < events for which plant is designed. Staff notes that a flood hazard reevaluation report (FHRR) was submitted for PSL Units 1 and 2 (ADAMS Accession No. ML15083A306; 2015). However, the application does not appear to include this report in its screening analysis.

With regards to external flooding, according to the FHRR, all flood causing mechanisms, except Local Intense Precipitation (LIP) and Probable Maximum Storm Surge (PMSS) associated with the Probable Maximum Hurricane (PMH), are bounded by the current licensing basis (CLB).

The NRC December 15, 2017, staff assessment of the PSL Flood Evaluation (FE) (ADAMS

ML17325B630) concluded that effective flood protection, if appropriately implemented, exists at PSL. Regarding Unit 2, portable stop logs are used for flood protection when directed by procedure. The NRC staff notes that Criterion C1 is provided within the context of the design basis and notes that the use of temporary barriers contingent on procedural compliance and operator action is usually not considered as part of the design basis. It is unclear to the NRC staff if the use of stop logs (as portable equipment) is allowed to be part of the licensing basis since it requires significant operator action.

a) Confirm the use of Unit 2 stop logs for flood protection is part of the PSL CLB.

b) If the Unit 2 stop logs are not part of the PSL CLB, then provide additional justification, such as a second screening criterion, that would allow the screening of external flooding when using procedurally directed installation of the stop logs as temporary barriers.

RESPONSE

a) St. Lucie Unit 2 UFSAR Section 2.4.2.2, Flood Design Consideration, discusses exterior doors and penetrations which lead to areas containing safety related equipment being made watertight through the use of boots, waterstops and waterproofing.

St. Lucie Unit 2 UFSAR Section 3.4.1, Flood Protection, refers to Figure 3.4-1 that shows details for waterproofing at penetrations and interconnections between seismic Category I structures. Penetrations for pipes or electrical ducts are either encased in concrete where they penetrate the wall or enclosed in a pipe boot where sleeves are used. Boots are not used below the normal groundwater table.

St. Lucie Unit 2 UFSAR Section 3.4.1 continues to state that based on probable maximum flood high water level, wave runup level and plant island elevation, flood protection stop logs at entrances whose minimum elevation is at least +19.5 feet to safety related buildings are not deemed necessary. Additional wave runup protection is provided to the entrances of the Fuel Handling Building and Reactor Auxiliary Building (RAB) by the presence of adjacent buildings and structures. Since no permanent structures are located on the south side of the Reactor Auxiliary Building, additional wave runup protection has been provided by installing stop logs in the entrance on the south wall and the southernmost entrance on the east wall.

Figure 3.4-6 details the stop logs provided for the two RAB openings. Rectangular aluminum stop logs would be stacked to Elevation 22.0 feet and secured with bolts.

Gaskets provide a seal at both the bottom and sides of the protected openings.

Additional security in this design is provided by the ability to bolt the stop logs against the opening frame thus assuring that vertical or horizontal separation cannot take place.

The stop logs are stored onsite in a manner that reserves their readiness for use. When a hurricane watch is posted for the plant, the stop logs are removed from storage and prepared for installation; with actual installation occurring when the hurricane warning is posted for the plant.

St. Lucie Unit 2 Technical Specification 3/4.7.6, Flood Protection, Limiting Condition for Operation 3.7.6.1 states, Flood protection shall be provided for the facility site via stoplogs which shall be installed on the southside of the RAB and the southernmost door on east wall whenever a hurricane warning for the plant is posted.

St. Lucie Unit 2 Technical Specification Bases for 3/4.7.6, Flood Protection, states, The limitation on flood protection ensures that facility protective actions will be taken in the event of flood conditions. The installation of the stoplogs ensures adequate protection for wave run-up effects where no permanent adjacent structures exist and provides protection to safety-related equipment. The maximum wave runup from the probable maximum flood (PMF) has been calculated to be elevation 18.0 feet Mean Low Water (MLW).

Information in the St. Lucie Unit 2 UFSAR, Technical Specifications, and Technical Specifications Bases confirm the use of Unit 2 stop logs for flood protection is part of the St. Lucie Unit 2 Current Licensing Bases.

b) As the St. Lucie Unit 2 stop logs are part of the St. Lucie Unit 2 Current Licensing Bases, no additional information is required to be provided.