L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For.

Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For.

ML19217A333
Person / Time
Site:	Prairie Island
Issue date:	08/05/2019
From:	Sharp S Northern States Power Company, Minnesota, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2018-LLA-0196, L-PI-19-029
Download: ML19217A333 (17)

Text

1717 Wakonade Drive Welch, MN 55089 August 5, 2019 L-PI-19-029 10 CFR 50.90 10 CFR 50.69 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0196)

References:

1) Letter (L-PI-18-012) from NSPM to the NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated July 20, 2018 (ADAMS Accession No. ML18204A393)

2) Email from the NRC to NSPM, Request for Additional Information RE:

Prairie Island 50.69 Amendment Request, dated February 26, 2019 (ADAMS Accession No. ML19057A165)

3) Letter (L-PI-19-014) from NSPM to the NRC, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0196), dated April 29, 2019 (ADAMS Accession No. ML19119A216)

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requested an amendment to adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors, for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) Report NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, as endorsed by Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance.

The NRC identified the need for additional information and provided the Request for Additional Information (RAI) in Reference 2 and NSPM responded to the RAI in Reference 3.

Document Control Desk Page 2 This supplement supersedes and completely replaces the following responses provided in Reference 3:

RAI 1 - Proposed License Condition RAI 6 - Key Assumptions and Uncertainties that Could Impact the Application RAI 7 - Dispositions of Possible Key Assumptions and Uncertainties RAI 8 - Flowserve N9000 [Reactor Coolant Pump] and Abeyance Seal Modeling

• to the Enclosure - Table A.1 - 10 CFR 50.69 Implementation Items The information provided in this letter does not alter the evaluations performed in accordance with 10 CFR 50.92 in Reference 1.

Please contact Mr. Peter Gohdes at (612) 330-6503 if additional information or clarification is required.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

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

Executed on August ~

, 2019.

PYw?

Scott Sharp Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

L-PI-19-029 NSPM Enclosure Page 1 of 13 Supplement to Request for Additional Information Response Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

1.0 BACKGROUND

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requested an amendment to adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors, for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) Report NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 2), as endorsed by Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (Reference 3). The NRC identified the need for additional information and provided the Request for Additional Information (RAI) in Reference 4 and NSPM responded to the RAI in Reference 5.

This supplement supersedes and completely replaces the following responses provided in Reference 5:

• RAI 1 - Proposed License Condition

• RAI 6 - Key Assumptions and Uncertainties that Could Impact the Application

• RAI 7 - Dispositions of Possible Key Assumptions and Uncertainties

• RAI 8 - Flowserve N9000 [Reactor Coolant Pump] and Abeyance Seal Modeling

• Attachment 1 to the Enclosure - Table A.1 - 10 CFR 50.69 Implementation Items 2.0 SUPPLEMENT TO REQUEST FOR ADDITIONAL INFORMATION RESPONSES RAI 1 - Proposed License Condition 10 CFR 50.69(b)(2)(ii) requires that a LAR to implement 50.69 include a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant(including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques) are adequate for the categorization of SSCs.

10 CFR 50.69(c)(1)(i) and (ii) 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.

L-PI-19-029 NSPM Enclosure Page 2 of 13 The guidance in NEI 00-04 allows licensees to implement different approaches, depending on the scope of their PRA (e.g., the approach if a seismic margins analyses is relied upon is different and more limiting than the approach if a seismic PRA is used). RG 1.201, Revision 1 states that as part of the NRC's review and approval of a licensee's or applicant's application requesting to implement 50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the licensee's categorization approach.

Section 3 of the LAR states that the licensee has submitted a separate LAR dated May 18, 2018 (ADAMS Accession No. ML18138A402), as supplemented on July 10, 2018 (ADAMS Accession No. ML18074A308), requesting revision of the license condition associated with implementation of NFPA 805. Table A-1 of this LAR, Table A Risk Significant Modifications Related to Implementation of NFPA 805 lists several risk-significant plant modifications that are credited in the fire PRA model but which are not yet installed in the plant. Because of the NRC staffs concurrent review of a separate LAR to revise the NFPA 805 modifications, the list provided in Table A-1 may yet change. Therefore, Table A-1 may not contain all modifications that would affect the plant PRA models. The NRC staff notes that the fire PRA model used for SSC categorization shall reflect the as-built, as-operated plant. This can be accomplished by completion of all NFPA 805 required modifications that affect PRA models or, if not all of these modifications are completed, by ensuring the as-built, as-operated plant PRA risk results satisfy all RG 1.174 acceptance guidelines.

Section 2.3 of the LAR Enclosure proposed the following license condition:

NSPM is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendment request dated July 20, 2018.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

NSPM shall complete the modifications listed in Table A-1 of the license amendment request dated July 20, 2018, prior to implementation.

The proposed license condition does not explicitly address the use of PRA and non-PRA approaches or provide assurance that PRA models used for SSC categorization reflect the asbuilt, as-operated plant. The final paragraph proposed in the license condition in the LAR references the modifications listed in Table A-1 of the LAR. RG 1.174 guidance is that the PRA used to support an application is technically acceptable and reflects the as-built, as-operated plant. The LAR provided insufficient information for the NRC staff to confirm that the sub-set of NFPA-805 modifications in Table A-1 were necessary and sufficient, and notes that additional method or plant changes may be required to yield a technically acceptable PRA that reflects the as-built, as-operated plants.

L-PI-19-029 NSPM Enclosure Page 3 of 13 Therefore, the staff has included a general statement in the final paragraph of the sample licensee condition provided below that is intended to ensure that all changes that are required to complete the transition to NFPA 805 and that are also modelled in the PRA are completed prior to implementation of the 50.69 categorization process.

Provide a license condition that explicitly addresses all the categorization approaches used by the staff and all the NFPA 805 changes that affect the PRA, e.g.:

NSPM is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding and fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

NSPM shall [ensure that the fire PRA model used for SSC categorization reflects the as-built, as-operated plant using the fire PRA and plant configuration that will be accepted to support final NFPA-805 implementation for both PINGP units at the time of the 50.69 categorization] prior to implementation.

Note that if additional implementation items are identified, the license condition may need to be expanded to address them.

NSPM Response to RAI 1 NSPM proposes the following license condition:

NSPM is approved to implement 10 CFR 50.69 using the approaches for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding and internal fire, with the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA

L-PI-19-029 NSPM Enclosure Page 4 of 13 evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds) updated using the external hazard screening significance criteria identified in ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in PINGP License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

NSPM will complete the implementation items listed in Attachment 1 of NSPMs letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69.

NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

RAI 6 - Key Assumptions and Uncertainties that Could Impact the Application 10 CFR 50.69(c)(1)(i) requires the licensee to consider the results and insights from the PRA during categorization. The guidance in NEI 00-04 specifies sensitivity studies to 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 importance of components. NEI 00-04 guidance states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

The guidance in NEI 00-04 specifies that sensitivity studies be conducted for each PRA model to address sources of 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 importance of components. NEI 00-04 guidance states that applicable sensitivity studies from characterization of PRA adequacy should be considered.

Section 3.2.7 of the LAR states that the detailed process of identifying, characterizing and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, March 2009 (Revision 0) (ADAMS Accession No. ML090970525) and Section 3.1.1 of EPRI Technical Report (TR)-1016737. The NRC staff notes that one of these sources has been superseded. Revision 1 of NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, dated March 2017 (ADAMS Accession No. ML17062A466) references updated EPRI guidance in TR-1026511, Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty (2012).

L-PI-19-029 NSPM Enclosure Page 5 of 13 The NRC staff notes that Stages C, D, E, and F of NUREG-1855 (Revision 1) provides guidance on how to identify key sources of uncertainty relevant to the application.

LAR Section 3.2.7 states that Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application, and are provided in Attachment 6. Attachment 6 of the LAR contains 10 key assumptions/sources of uncertainties. The LAR does not describe how the key assumptions and sources of uncertainty were identified, and whether the outcome described in the LAR was the result of a comprehensive examination for key assumptions and sources of uncertainty using recent industry guidance.

i. Provide a description of the process used to determine how the candidate key assumptions and sources of uncertainty were identified and evaluated for the internal event (including internal flooding) and fire PRAs. Include in the discussion explanation of how uncertainty issues associated with plant specific features, modeling choices, and generic industry concerns were addressed. Also, include in the description explanation of whether the assumptions and sources of uncertainty documented in the PRA modeling notebooks were reviewed to determine if they could have a possible impact on the application.

ii. Describe how the process described in Part (i) above is consistent with NUREG-1855, Revision 1, or another NRC-accepted method.

iii. If the process of identifying key assumption or sources of uncertainty for these PRA models cannot be justified, provide the results of an updated assessment of key sources of uncertainty or assumptions. Include a description of the specific assumptions and sources of uncertainty key to this application in enough detail so that its impact on the application can be clearly understood and a specific sensitivity study could be defined to examine the impact on 50.69 categorization.

iv. If the response to part (iii) above results in the identification of key assumptions or sources of uncertainty that should be addressed as part of the 50.69 categorization then propose a mechanism to ensure that the identified sensitivity study is performed as part of PINGPs 50.69 categorization process.

NSPM Response to RAI 6.i The sources of uncertainty evaluation for the internal events PRA considers both plant-specific sources of uncertainty and the generic uncertainties identified in EPRI TR-1016737 (Reference 6). At the time of the LAR submittal, the fire PRA considered the plant-specific uncertainty sources, but did not specifically address the EPRI generic sources as provided in EPRI TR-1026511 (Reference 7). However, a subset of these generic uncertainties were considered within the context of the plant-specific uncertainty evaluation. Both modeling uncertainty and completeness uncertainty issues were examined for both PRAs.

L-PI-19-029 NSPM Enclosure Page 6 of 13 The internal events PRA includes an evaluation of the sources of uncertainty. All identified sources of uncertainty were compiled and characterized in the Uncertainty Notebook by reviewing modeling assumptions from all of the PINGP PRA technical element notebooks for the base case models using an approach that is consistent with the ASME/ANS RA-Sa-2009 (Reference 8) requirements for identification and characterization of uncertainties and assumptions. This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1 (Reference 9).

The fire PRA sources of uncertainty evaluation has been similarly updated to compile and characterize plant-specific assumptions and associated sources of model uncertainty as well as the generic sources of uncertainty presented in EPRI TR-1026511 based on the most recent fire model update. This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1.

To assess the impact of sources of uncertainties on 10 CFR 50.69 system categorizations, a review of the base case sources of uncertainty for the internal events and fire PRAs was performed prior to the submittal of the original LAR and has now been updated based on the latest PRA sources of uncertainty evaluations. Each identified uncertainty was evaluated with respect to its potential to significantly impact the risk ranking evaluations that will be performed to support the categorization effort. Previously identified sources of uncertainty for internal events were investigated further and most were determined to have negligible impacts on the 50.69 process. See the responses to RAI 7 concerning the uncertainties that were removed. A single key uncertainty remains that could impact the 50.69 categorization process. Similarly, the updated fire PRA uncertainty review against the current fire PRA model determined that there are no relevant sources of uncertainty that pertain to the use of the fire PRA for the 50.69 categorization process. This evaluation meets the intent of the screening portion of steps C-2 and E-2 of NUREG-1855, Revision 1.

Based on the updated evaluation, Attachment 6 of the LAR is replaced in its entirely with the following:

L-PI-19-029 NSPM Enclosure Page 7 of 13 : Disposition of Key Assumptions/Sources of Uncertainty Assumption/Uncertainty Discussion Disposition Internal Events PRA Model Key Assumptions/Sources of Uncertainty Basis for Human Error Probabilities (EPRI-identified generic source of modeling uncertainty)

Human Reliability Analysis (HRA) is a continually evolving discipline. The human error probabilities were obtained using the current EPRI HRA calculator consistent with the Fire HRA Methodology described in NUREG-1921 (Reference 10).

The internal events human error probabilities were obtained using guidance from NUREG/CR-1278 (Reference 11), and NUREG/CR4772 (Reference 12).

Given the methodologies used, the impact of any remaining uncertainties is expected to be small.

The PINGP PRA model is based on industry consensus modeling approaches for its HEP calculations, so this is not considered a significant source of uncertainty.

As directed by NEI 00-04, human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled Human Error Probabilities are accounted for in the 50.69 application.

Fire PRA Model Key Assumptions/Sources of Uncertainty None Identified NSPM Response to RAI 6.ii As discussed in the response to part i of RAI 6 above, the approach used to identify the sources of uncertainty for consideration meets the intent of NUREG-1855, Revision 1 for the steps identified (C-1, C-2, E-1, and E-2).

The ultimate goal in assessing model uncertainty is to determine whether (and the degree to which) the risk metric results challenge or exceed the quantitative acceptance guidelines for the application due to sources of model uncertainty and related assumptions. For 10 CFR 50.69 categorization, the PRA acceptance guidelines are threshold values for FussellVesely (FV) and Risk Achievement Worth (RAW) for each SSC being categorized, above which the SSC is categorized as candidate high safety significant (HSS) and below which the SSC is categorized as candidate low safety significant (LSS). As described in Step E-2 of the NUREG, each relevant uncertainty/assumption requires some sort of sensitivity analysis, and each sensitivity performed to evaluate an uncertainty/assumption involves some change to the PRA results. Since any change to the PRA results has the potential to change the FV and RAW importance measures for all components (SSC), every relevant

L-PI-19-029 NSPM Enclosure Page 8 of 13 uncertainty/assumption has the potential to challenge the acceptance guidelines. That is, since RAW and FV are relative importance measures, any change to any part of the model will generate a new set of cutsets and potentially impact the RAW and FV for every SSC. Thus, the only way to evaluate the impact of a sensitivity is to quantify the sensitivity case and compare the FV and RAW values for all SSCs against the base case FV and RAW values to determine if any exceed the HSS threshold in the sensitivity case that did not previously do so.

NSPM Response to RAI 6.iii As documented in the responses to parts i and ii of RAI 6, the approach used to identify significant sources of uncertainty for the 10 CFR 50.69 application and to evaluate their impacts is adequate and appropriate to ensure that these sources of uncertainty will not impact the system categorization evaluations for SSC risk significance.

NSPM Response to RAI 6.iv This question is not applicable to PINGP based on the information provided in response to parts i, ii, and iii of RAI 6.

RAI 7 - Dispositions of Possible Key Assumptions and Uncertainties 10 CFR 50.69(c)(1)(i) and (ii) 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 sensitivity studies to 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 importance of components. NEI 00-04 guidance states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

The dispositions to the 10 assumption/uncertainty items identified in Attachment 6 of the LAR can be summarized as follows: (1) four state that sensitivity studies will be performed as necessary, (2) four (including all three for the fire PRA model) conclude that it is not a source of uncertainty for the application, (3) one is addressed by the general NEI 04-10 human error probability (HEP) sensitivity analysis, and (4) one will be treated on a case-by-case basis as needed.

i.

Explain how a sensitivity study will be determined to be necessary and as needed.

ii.

The sixth uncertainty item in Attachment 6 concerns thermally-induced steam generator tube rupture (TI-SGTR). The LAR states that TI-SGTR is primarily a phenomenological uncertainty for Large Early Release Frequency (LERF) and that its impact on LERF is

L-PI-19-029 NSPM Enclosure Page 9 of 13 low enough such that no impact on 50.69 categorization is expected. However, the LAR also states that TI-SGTR can be significant for some non-station black out sequences.

a. Provide justification, such as a sensitivity study, that the exclusion of a TI-SGTR does not impact 50.69 categorization for any SSCs.

b. Alternatively, propose a mechanism to ensure this issue is addressed as a sensitivity study during the 50.69 categorization process.

NSPM Response to RAI 7.i The specific uncertainties dispositioned in Attachment 6 of the LAR as requiring sensitivity studies as needed are primarily items involving exclusion of low probability failure modes in the modeling of specific plant systems, such as Residual Heat Removal (RHR), Main Feedwater, Circulating Water, etc. Additional evaluations were performed to determine whether these listed uncertainties could be excluded. For uncertainties that could not be excluded, studies were performed to determine the impact of these uncertainties on risk rankings of affected components.

The additional evaluations found that the listed uncertainties associated with the RHR and Main Feedwater systems modeling could be excluded based on screening criteria specified in ASME/ANS PRA Standard RA-Sa-2009. Therefore, no additional sensitivity evaluations will be necessary for these uncertainties.

The uncertainty associated with failure to trip the Circulating Water pumps was evaluated using a sensitivity study. The sensitivity study added the operator action to the PRA model to determine the impact of this additional action on risk ranking of components. The results of the study identified no events across both units that could transition from LSS to HSS on the basis of the risk results that would not already be identified as HSS as a result of the base importances or the NEI 00-04 required sensitivity studies. Therefore, no additional sensitivity evaluations will be necessary for this uncertainty.

The treatment of uncertainties pertaining to thermally-induced steam generator tube rupture (SGTR) is addressed in the response to RAI 7 part ii.

The uncertainty associated with the assumption that low pressure RHR piping will always rupture upon exposure to Reactor Coolant System (RCS) pressure, resulting in an interfacing system loss of coolant accident (ISLOCA) event, is a conservative treatment of ISLOCA impacts. ISLOCA is a small contributor to CDF and a key contributor for LERF. Reducing the probability of pipe rupture given an ISLOCA would reduce the importance of the ISLOCA contributors and possibly increase the risk significance of other SSCs, especially for LERF. A sensitivity study was performed that reduced the probability of a low pressure pipe rupture on exposure to RCS pressure from 1.0 to 0.05. The results of the study identified no events across both units that could transition from LSS to HSS on the basis of the risk results that would not already be identified as HSS as a result of the base importances or the NEI 0004 required sensitivity studies. Therefore, no additional sensitivity evaluations will be necessary.

L-PI-19-029 NSPM Enclosure Page 10 of 13 NSPM Response to RAI 7.ii.a For the thermally-induced failure of a steam generator tube, the PINGP PRA model follows WCAP-16341-P, Simplified Level 2 Modeling Guidelines (Reference 13), which provides a basis for updated Level 2 analyses. This WCAP provides a common, standardized method for PWRs with large dry containments. The guidance particularly addresses the latest understanding for thermallyinduced steam generator tube ruptures and other Level 2 issues.

The text included in the Discussion column of Attachment 6 of the LAR was derived from the overall assessment of all base PRA sources of uncertainty that is documented in the internal events PRA uncertainty notebook. The text that was developed for that notebook was specific to applications with a focus on SGTR and is not applicable to 10 CFR 50.69. As noted in the Disposition column of Attachment 6 of the LAR, this was judged to not represent a key source of uncertainty in the 50.69 application since CDF would not be impacted and the overall impacts on LERF were expected to be small.

Since the PINGP PRA uses a TI-SGTR modeling approach that is consistent with recent industry approaches and is appropriate for determination of LERF, this does not represent a key source of uncertainty in the 10 CFR 50.69 application.

NSPM Response to RAI 7.ii.b As discussed in the RAI 7.ii.a response, the TI-SGTR modeling treatment is consistent with industry methods, therefore, no additional mechanism is needed to address this issue.

RAI 8 - Flowserve N9000 RCP and Abeyance Seal Modeling 10 CFR 50.69(c)(1)(i) and (ii) 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 sensitivity studies to 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 importance of components. NEI 00-04 guidance states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

Section 3.3 of the LAR states a focused-scope peer review was conducted to address the incorporation of Flowserve N9000 Reactor Coolant Pump seals. The disposition to F&O SYA1701 in Attachment 3 of the LAR states, the N-9000 RCP seal model must obtain NRC review and approval. The NRC staff notes that the N-9000 RCP seal model is approved for Combustion Engineering (CE) plants using the guidance of WCAP-16175-P-A with conditions,

L-PI-19-029 NSPM Enclosure Page 11 of 13 limitations, and modifications in the NRC staff safety evaluation (SE) (ADAMS Accession No. ML071130383). The staffs SE that [a]dditional conditions, limitations, and modifications are provided in this SE to address some of the issues that must be addressed by application of TR WCAP-16175-P, Revision 0, RCP seal failure model to non-CE plants. The LAR did not address if the PINGP PRA model implementation used all applicable guidance in this WCAP.

The NRC staff also notes that abeyance seals are sometimes used as a backup to Flowserve RCP seal packages to limit leakage if excessive flow from the mechanical face seals occurs (ADAMS Accession No. ML15222A357). There is currently no NRC accepted methodology to model the abeyance seal in PRAs.

In light of these observations:

i. Confirm that the PRA model implementation of the N-9000 RCP seal was in accordance with WCAP-16175-P-A and addressed all applicable NRC staff conditions, limitations, and modifications as described in the associated safety evaluation. Alternatively, describe the methodology used to model the N-9000 RCP seal and justify that this methodology is acceptable.

ii. If the baseline PRA model of record used for this LAR credits an abeyance RCP seal, provide the PRA methodology to model the abeyance seal and describe how this inclusion impacts the categorization.

iii. Propose a mechanism that ensures an NRC approved abeyance RCP seal model is available before incorporation of an abeyance seal into the PRA MORs.

NSPM Response to RAI 8.i NSPM has completed a review of the PINGP PRA model implementation of the N-9000 RCP seal and confirmed that it was completed in accordance with WCAP-16175-P-A, Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants, (see Reference 14 for non-proprietary version) and addressed all applicable NRC conditions, limitations, and modifications as described in the associated safety evaluation.

NSPM Response to RAI 8.ii The NSPM PRA model used for categorization will only credit the abeyance RCP seal after the NRC has approved the modeling approach for the abeyance seal. Prior to modeling approach approval, the PRA model used for categorization will not credit the abeyance seal.

NSPM Response to RAI 8.iii The proposed mechanism is in the response to RAI 1.

L-PI-19-029 NSPM Enclosure Page 12 of 13

3.0 REFERENCES

1.

Letter (L-PI-18-012) from NSPM to the NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated July 20, 2018 (ADAMS Accession No. ML18204A393)

2.

NEI Guideline 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005 (ADAMS Accession No. ML052910035)

3.

NRC Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, dated May 2006 (ADAMS Accession No. ML061090627)

4.

Email from the NRC to NSPM, Request for Additional Information RE: Prairie Island 50.69 Amendment Request, dated February 26, 2019 (ADAMS Accession No. ML19057A165)

5.

Letter (L-PI-19-014) from NSPM to the NRC, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0196), dated April 29, 2019 (ADAMS Accession No. ML19119A216)

6.

EPRI Technical Report TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, dated December 2008

7.

EPRI Technical Report TR-1026511, Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, dated December 2012

8.

ASME 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, dated February 2, 2009

9.

NRC NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)

10.

NRC NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, dated July 2012 (ADAMS Accession No. ML12216A104)

11.

NUREG/CR-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, dated August 1983 (ADAMS Accession No. ML071210299)

12.

NUREG/CR-4772, Accident Sequence Evaluation Program Human Reliability Analysis Procedure, dated February 1987

L-PI-19-029 NSPM Enclosure Page 13 of 13

13.

WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, dated November 2005

14.

Westinghouse Electric Company WCAP-16175-NP-A, Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants, Revision 0, dated March 2007 (ADAMS Accession No. ML071130383)

ENCLOSURE, ATTACHMENT 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 Supplement to Response to Request for Additional Information:

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors 10 CFR 50.69 IMPLEMENTATION ITEM (1 Page Follows)

L-PI-19-029 NSPM Enclosure, Attachment 1 Page 1 of 1 Table A.1 - 10 CFR 50.69 Implementation Items No.

Implementation Item

1.

The NSPM PRA model used for categorization will only credit the abeyance RCP seal after the NRC has approved the modeling approach for the abeyance seal. Prior to modeling approach approval, the PRA model used for categorization will not credit the abeyance seal.