ML24366A003

NRR E-mail Capture - Final RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Adoption of 10 CFR 50.69

ML24366A003
Person / Time
Site:	Cook
Issue date:	12/30/2024
From:	Scott Wall NRC/NRR/DORL/LPL3
To:	Scarpello M Indiana Michigan Power Co
Wall S
References
L-2024-LLA-0025
Download: ML24366A003 (11)

Text

From:

Scott Wall Sent:

Monday, December 30, 2024 3:23 PM To:

Michael K. Scarpello Cc:

Helen L Levendosky; Joe Tanko

Subject:

FINAL RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Adoption of 10 CFR 50.69 (EPID No. L-2024-LLA-0025)

Dear Michael Scarpello,

By letter dated March 6, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24073A234), Indiana Michigan Power Company (I&M, the licensee) submitted a license amendment request (LAR) for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP). The proposed LAR would adopt Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, into the licensing basis for CNP. The provisions of 10 CFR 50.69 allow licensees to use an integrated, systematic, risk-informed process for categorizing structures, systems, and components (SSCs) according to their safety significance.

A licensee that has adopted 10 CFR 50.69 may specify alternative treatments for SSCs that have low safety significance.

The NRC staff has reviewed the submittal and determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). On December 30, 2024, the I&M staff indicated that a response to the RAIs would be provided by February 7, 2025.

If you have questions, please contact me at 301-415-2855 or via e-mail at Scott.Wall@nrc.gov.

Scott P. Wall Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301.415.2855 Scott.Wall@nrc.gov Docket Nos. 50-315 and 50-316

Enclosure:

Request for Additional Information cc: Listserv RAI (10 CFR 50.69)

REQUEST FOR ADDITIONAL INFORMATION

LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 INTRODUCTION By letter dated March 6, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24073A234), Indiana Michigan Power Company (I&M, the licensee) submitted a license amendment request (LAR) for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP). The proposed LAR would adopt Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, into the licensing basis for CNP.

The provisions of 10 CFR 50.69 allow licensees to use an integrated, systematic, risk-informed process for categorizing structures, systems, and components (SSCs) according to their safety significance. A licensee that has adopted 10 CFR 50.69 may specify alternative treatments for SSCs that have low safety significance.

The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing the application and has determined that the following additional information is required in order to complete the review.

Probabilistic Risk Assessment Licensing Branch A (APLA) Questions Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (ML061090627), endorses, with regulatory positions and clarifications, the Nuclear Energy Institute (NEI) guidance document NEI 00-04, Revision 0, 10 CFR 50.69 SSC [Structure, System, and Component] Categorization Guideline (ML052910035), as one acceptable method for use in complying with the requirements in 10 CFR 50.69. Section 3.1.1 of the LAR dated March 6, 2024, states that I&M will implement the risk categorization process of 10 CFR 50.69 in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201.

The following questions and topics for discussion are intended to help the NRC staff determine if the licensee has implemented the guidance appropriately in NEI 00-04, as endorsed by RG 1.201, as a means to demonstrate compliance with all of the requirements in 10 CFR 50.69, including acceptability of the PRA models.

APLA-RAI-1 (Audit Question 2) - Peer Review Facts and Observations (F&Os) Findings In 10 CFR 50.69(c)(1)(i) and (ii), the regulations require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process and 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. Revision 2 of RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML090410014), and Revision 3 of RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML20238B871) both describe acceptable approaches for determining the acceptability of a base PRA used in regulatory decision-making for commercial light-water nuclear power plants. Both revisions of RG 1.200 endorse, with clarifications, the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) PRA Standard ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. Section 3.3 of the LAR states that the PRA models have been assessed against RG 1.200, Revision 2.

As discussed in RG 1.200, it is recognized that a PRA may not satisfy each technical requirement to the same degree (i.e., capability category as used in the ASME/ANS PRA standard); that is, the capability category achieved for the different technical requirements may vary. This variation can range from (1) the minimum needed to meet the characteristics and attributes for each technical element to (2) the minimum to meet current good practice (i.e.,

state-of-practice) for each technical element. Further, which capability category is needed to be met for each supporting requirement is dependent on the specific application. In general, the NRC staff considers Capability Category II (CC-II) of the ASME/ANS PRA standard to provide a level of detail that is acceptable for the majority of applications. However, for some applications, Capability Category I (CC-I) may be acceptable for some supporting requirements.

a. [Capability Category] Please confirm for all peer reviews and independent closure reviews of facts and observations (F&Os) for the Full Power Internal Events PRA and internal floods PRA, that: (1) all supporting requirements (SRs) met at CC-I have been identified in the table in Attachment 3 of Enclosure 2 of the LAR; and (2) that no SRs were determined to be Not met. Alternatively, provide a list of all SRs that are CC-I and Not met and a disposition for why this is acceptable for the application (i.e., risk-informed categorization of SSCs in accordance with 10 CFR 50.69).

b. [F&O 4-4] In order to support risk-informed categorization, PRA models need to reflect the as-built, as-operated plant. In Attachment 3 of Enclosure 2 of the LAR, the disposition for F&O 4-4 states, in part: Some walkdowns and interviews have been performed and did not identify any necessary modeling changes, the same outcome is expected for those systems that still need walkdowns and interviews performed. Please identify which systems still need walkdowns/interviews. Provide additional justification as to why these walkdowns/interviews are not expected to result in changes to the PRA model that could impact the application (i.e., risk-informed categorization of SSCs in accordance with 10 CFR 50.69). Alternatively, propose a mechanism to ensure that the applicable walkdowns/interviews are performed prior to implementation of 10 CFR 50.69.

c. [F&O 2-19 and FQ-D1-02] In Attachment 3 of Enclosure 2 of the LAR, the F&O Description for F&O 2-19 states that the Large Early Release Frequency (LERF) analysis uses conservative assessments, and these portions do not meet CC-II. One example of conservatism is the assumption that a steam generator tube rupture (SGTR) is a containment bypass event without considering success of secondary side isolation. The associated disposition states that pressure and temperature induced SGTR events are modeled as progressing directly to LERF and that this ensures an over-estimation. The discussion of Fire PRA F&O FQ-D1-02 also refers to the disposition of F&O 2-19.

As discussed in NUREG-1855, Rev. 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (ML17062A466), Section 7.2.3.3, the use of conservative assumptions in one part of the model can mask the significance of another part of the model (i.e., the importance measures of other modeled SSCs) and the impact of this conservative bias should be assessed. The dispositions provided for F&Os 2-19 and FQ-D1-02 are not specific to the impact on risk-informed categorization and do not address the possibility of masking risk insights due to conservative bias that may lead to the inappropriate categorization of SSCs. The guidance in NEI 00-04 also recommends the use of use of sensitivity studies identified in the characterization of PRA adequacy, if applicable. Considering these observations, please provide:

1. Qualitative or quantitative justification (e.g., sensitivity studies) that supports the determination that conservative modeling assumptions that impact LERF results (e.g., not modeling secondary side isolation for a SGTR) will not adversely impact the categorization of SSCs per the 10 CFR 50.69 application and is not a key assumption/source of uncertainty for the application. If determined to be a key assumption/source of uncertainty, please provide updates to Attachment 6 and Section 3.2.7 of the LAR, as needed.

OR

2. Propose a mechanism to ensure that the issues identified in the F&Os are resolved prior to implementation of 10 CFR 50.69. For example, the model could be updated, and an independent F&O closure assessment could be performed. (Note: If any proposed PRA model changes constitute an upgrade to the PRA model as defined in the PRA Standard ASME/ANS RA-Sa-2009, a focused-scope peer review would be required which could result in additional F&Os that need to be addressed prior to implementing the SSC categorization process.)

d. [F&O 2-4] In Attachment 3 of Enclosure 2 of the LAR, the F&O Description for F&O 2-4 states that all containment failures caused by hydrogen combustion are assumed to contribute to LERF. The associated disposition states that the current implementation is conservative and that some sequences could potentially be screened out based on containment failure location, but that improvements in modeling would not result in a significant improvement in the overall realism of the PRA results. It is unclear to the NRC staff how significant the contribution from hydrogen combustion scenarios is to LERF results.

Please provide qualitative or quantitative justification that supports the determination that conservative modeling assumptions related to containment failure due to hydrogen combustion will not adversely impact the categorization of SSCs per the 10 CFR 50.69 application and is not a key assumption/source of uncertainty for the application. If determined to be a key assumption/source of uncertainty, please provide updates to and Section 3.2.7 of the LAR, as needed.

e. Section 3.1.1 of the LAR notes that previous versions of the following PRA models were submitted and accepted by the NRC for the listed applications. Further in Section 3.2.1 of the LAR the licensee states in part, "[t]he CNP categorization process for the internal events and internal flooding hazards will use a technically acceptable and independently peer reviewed plant-specific PRA model." It is unclear to the NRC staff if the FPIE 2023-R0 PRA model to be used in the categorization process is the same base model as for Revision

15MORW R1. Please confirm that FPIE 2023-R0 is the same base model that was full-scope peer reviewed in July 2015 as stated in Section 3.3 of the LAR.

APLA-RAI-2 (Audit Question 4) - Passive Categorization In Section 3.1.2 of the LAR, for the passive categorization process, the licensee states, in part:

Component supports, if categorized, are assigned safety based upon one of the following approaches:

A combination of restraints or supports such that the [Low Safety Significant (LSS)] piping and associated SSCs attached to the [high safety significance (HSS)] piping are included in scope up to a boundary point that encompasses at least two supports in each of three orthogonal directions [27, 28].

It appears the references cited are not applicable to the LAR excerpt above. Please confirm References 27 and 28 and update appropriately.

APLA-RAI-3 (Audit Question 6) - Total Risk Considerations Revision 3 of RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ML17317A2560), provides risk acceptance guidelines in terms of change-in-risk in combination with total core damage frequency (CDF) and LERF. The guidance in NEI 00-04 includes an overall risk sensitivity study for all the LSS components to assure that if the unreliability of the components was increased, the increase in risk would be small (i.e., meet the acceptance guidelines of RG 1.174).

The guidance in RG 1.174 and Section 6.4 of NUREG-1855, Revision 1 states in part, for a Capability Category II risk evaluation, that the mean values of the risk metrics (total and incremental values) should be compared against the risk acceptance guidelines in RG 1.174.

The mean values referred to are the means of the probability distributions that result from the propagation of the uncertainties for the PRA input parameters and model uncertainties explicitly represented in the PRA models. In general, the point estimate CDF and LERF obtained by quantification of the cutset probabilities using mean values for each basic event probability does not produce a true mean of the CDF/LERF.

The LAR does not state whether the total CDF and LERF values presented in Attachment 2 of of the LAR are mean values. The NRC staff notes there is a small margin between the RG 1.174 LERF threshold of 1E-05 per year for LERF and the results for Units 1 and 2 of 9.91E-06 and 9.67E-06 per year, respectively, provided in the LAR, Attachment 2 of. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the state of knowledge correlation (SOKC) is unimportant (e.g., the risk results are well below the acceptance guidelines). Accordingly, the risk increase due to consideration of the SOKC could impact the conclusions of the NEI 00-04 Section 8 overall sensitivity study results by increasing the total LERF values above 1E-05 per year for CNP Units 1 and 2.

Considering the observations above, please address the following:

a. Provide a summary of how the SOKC investigation was performed for the base CNP PRA models used to support the 50.69 applications. In the summary, explain which fire PRA

parameters are assumed to be correlated beyond the component failure modes frequencies.

Include justification that consideration of the identified parameters is sufficient to estimate the impact of the SOKC on fire risk.

b. Demonstrate that for the NEI 00-04 Section 8 overall sensitivity study results, CNP will be in conformance with the RG 1.174 risk acceptance guidance after the Full Power Internal Events (including flooding) and Fire PRA models are updated to include the increase associated with SOKC (if needed).

Probabilistic Risk Assessment Licensing Branch C (APLC) Questions APLC-RAI-1 (Audit Question 1) - Integrated PRA Hazards Model 10 CFR 50.69(b)(2)(i) requires a licensee voluntarily choosing to implement 50.69 to apply for a license amendment with a description of the process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs. Paragraph (c)(1)(ii) of 10 CFR 50.69 requires that the SSC functional importance be determined using an integrated, systematic process. NEI 00-04, Section 5.6, Integral Assessment, discusses the need for an integrated computation using available importance measures. It further states that the integrated importance measure essentially weighs the importance from each risk contributor (e.g., internal events, fire, seismic PRAs) by the fraction of the total core damage frequency (or large early release frequency) contributed by that contributor. The guidance provides formulas to compute the integrated Fussell-Vesely importance (FV) and integrated Risk Achievement Worth (RAW).

Based on the information provided in the LAR, it is not clear to the NRC staff how the licensee proposes to address the integration of importance measures across all hazards (i.e., internal events, fire, and seismic).

Please address the following:

a. Explain how the integration of importance measures across hazards for the 10 CFR 50.69 categorization process will be performed.

b. Discuss how the importance measures for the PRA models (e.g., FV and RAW) are derived and justify why the importance measures generated do not deviate from the NEI guidance or Table 3-1 of the LAR. If the practice or method used to generate the integrated importance measures is determined to deviate from the NEI guidance, then provide justification to support why the integrated importance measures computed are appropriate for use in the categorization process.

c. Describe how the importance measures for the seismic PRA (e.g., FV and RAW) are derived considering that the seismic hazard is discretized into bins. The discussion should include how the same basic events, which were discretized by binning during the development of the seismic PRA, are then combined (i.e., combined across bins as well as across failure modes such as seismic and random failure modes) to develop representative importance measures. Further, discuss how they are compared to the importance measure thresholds in NEI 00-04. Provide justification to support the determined impact on the categorization results and describe how the approach is consistent with the guidance in NEI 00-04.

d. In the context of the integral assessment described in NEI 00-04, Section 5.6, it is understood that importance evaluations performed in accordance with the process in NEI 00-04 are determined on a component basis. However, the LAR and NEI 00-04 guidance does not make clear how the integrated importance measures are calculated for certain components. Specifically, in the seismic PRA, basic events that represent different failure modes for a component may not align with basic events in other PRA models.

Examples of such basic events include those that are specific to the seismic PRA (including implicitly modeled components) or basic events that represent a subcomponent modeled within the boundary of a component in the internal events PRA.

APLC-RAI-2 (Audit Question 2) - 2013 PRA Standard 10 CFR 50.69(c)(1)(i) requires the PRA to be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

In Section 3.3 of the LAR, an excerpt from PWROG-18062-P refers to an effective endorsement of Part 5 of Addendum B of the ASME/ANS RA-SB-2013 PRA standard for seismic modeling.

The NRC staff accepted ASME/ANS RA-S Case 1, Case for ASME/ANS RA-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment of Nuclear Power Plant Applications, dated November 22, 2017, for use in licensing actions on an interim basis in letter dated March 12, 2018 (ML18017A964). Further, in Revision 3 of RG 1.200, Appendix B endorses (with clarifications and qualifications) ASME/ANS RA-S Case 1, Case for ASME/ANS RA-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment of Nuclear Power Plant Applications, dated June 2020.

a. Given available regulatory basis for applicability of Code Case 1 for licensing actions, explain the purpose of DC Cooks statements about effective endorsement from a different context.

b. Confirm that the seismic PRA supporting the LAR addresses the staffs clarifications and qualifications on Code Case 1 as provided in the enclosure to the letter dated March 12, 2018 (ML18017A966) and included in Revision 3 of RG 1.200.

APLC-RAI-3 (Audit Question 3) - Open or Partially Resolved F&Os (Seismic PRA) 10 CFR 50.69(c)(1)(i) requires the PRA to be of sufficient quality and level of detail to support the categorization process and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

In Attachment 3 of Enclosure 2 of the LAR, there are numerous open and partially resolved findings from prior peer reviews of the seismic PRA. The following questions are related to those findings and the possible impact to SSC categorization under 10 CFR 50.69.

a. Seismic F&O 20-7 (later incorporated into F&O 1-1) states that more than one method should be considered to evaluate soil liquification to address epistemic uncertainty.

Provide the evaluation of the liquification hazard using an alternate method. If such an evaluation has not been performed, demonstrate that the open finding level F&O does not impact SSC categorization under 10 CFR 50.69.

b. Seismic F&O 20-7 (later incorporated into F&O 1-1) requested the evaluation of the lateral spreading hazard at the site. The original F&O states that Figures 6-7 and 6-9 of DC-COOK-PR-09 indicates there is a continuous layer of potentially liquefiable soils (in direction towards the lake) on borings B120, B124, B133, B142, and B141 between elevations of about 560 and 555 ft. I&M in contrast has stated that they consider Figure 6-7 to show no lateral continuity of the liquefiable boreholes.

Explain in detail why I&M has reached a different conclusion from the Seismic Peer Review Team for lateral spreading. Additionally, demonstrate that the open finding level F&O does not impact SSC categorization under 10 CFR 50.69.

c. Seismic F&O 2-1 remains open because the potential for cracking in the auxiliary building has not been fully dispositioned. The conservative or non-conservative impacts on the seismic PRA results has the potential to impact overall risk insights as they apply to the application of 50.69.

Demonstrate that the open finding level F&O does not impact SSC categorization under 10 CFR 50.69.

d. Seismic F&O 28-2 identified gross simplifications including modeling of the nuclear steam supply system (NSSS) equipment as lumped masses only (without inclusion of stiffness).

Some recommendations were made to close the F&O by either performing a sensitivity study to evaluate the impact of NSSS stiffness on the overall structure or correctly implementing ASCE 4-16, Section 3.7 criteria.

Provide a sensitivity study to evaluate the impact of NSSS stiffness on the overall structure and seismic PRA. Alternatively, demonstrate that the open finding level F&O does not impact SSC categorization under 10 CFR 50.69.

e. Seismic F&O 28-11 identified that structural variability is ignored when determining the seismic response of SSCs. In I&Ms disposition of the F&O, a statement is made that licensee staff will review the small number of risk-significant components on a case-by-case basis, adjusting the frequency range of interest (FROI) by an additional +/- 15 percent to ensure structural variability is captured in the fragility calculations. NEI 00-04 Rev. 0 requires sensitivity studies for seismic PRAs (Table 5-4), including any applicable sensitivity studies identified in the characterization of PRA adequacy.

Provide documentation of the referenced FROI adjustments. If the FROI adjustments have not been performed, will the adjustments be made as part of the 50.69 SSC categorization process?

f. Seismic F&O 28-13 states that the gap in power spectral density (PSD) as described in the F&O should be addressed per latest fragility guidance document. In I&Ms disposition of the F&O, I&M indicates that there are not significant gaps in energy near frequencies that are important to risk-significant fragilities based on a review of the non-interpolated PSDs and PSDs developed using a linear frequency interpolation.

Provide a graphical representation of the non-interpolated PSDs and PSDs developed using a linear frequency interpolation that supports the I&Ms position that no significant PSD gaps exist near the frequencies important to risk-significant fragilities. Confirm that the ground motion time histories used in fragility calculations are adequate for SSC categorization under 10 CFR 50.69.

APLC-RAI-4 (Audit Question 4) - External Hazard Screening 10 CFR 50.69(c)(1)(ii) requires an applicant while performing an SSC Categorization Process to determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant-specific PRA 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 following questions are related to the screening of external hazards as documented in the LAR Attachment 4 of Enclosure 2.

a. As part of the LAR, a variety of external hazards were screened out of consideration for the 10 CFR 50.69 categorization process in Attachment 4 of Enclosure 2. For the external hazard of ice cover - portions of the circulating water deicing system were credited as HSS to justify the exclusion of this hazard from the categorization process. Screening Criterion C1 (event damage potential is less than events for which the plant is designed), C4 (event is included in the definition of another event), and C5 (event develops slowly, allowing adequate time to eliminate or mitigate the threat) are all referenced as partial bases for screening the ice cover hazard.

With respect to the screening of the ice cover external hazard - please provide one of the following:

1. Justify why categorizing the proposed portions of the circulating water deicing system as HSS is adequate for the screening of this hazard consistent with NEI 00-04, Figure 5-6, Other External Hazards, and the discussion that if a component is credited as part of the safe shutdown paths evaluated, it is considered safety significant.

2. Provide a different technical basis and/or criterion for the screening this hazard.

b. [Release of Chemicals from On-site Storage] In order to screen out the external hazard of a release of chemicals from on-site storage, CA-03-01 documents an analysis showing a hydrazine spill at a storage location would not impact control room habitability.

Has a hydrazine spill been evaluated for locations during transit or use within the power block - not just storage? If not, justify why performance of such an evaluation would not impact SSC categorization under 10 CFR 50.69.

c. [Turbine-Generated Missiles] In order to screen out the external hazard of turbine-generated missiles, a turbine missile probability analysis was performed for the Unit 1 and Unit 2 Alstom low-pressure turbines.

Describe whether a similar analysis was done for any high-pressure turbines on-site at CNP.

If not, justify why the performance of such an analysis would not impact SSC classification under 10 CFR 50.69.

Hearing Identifier:

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Subject:

FINAL RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Adoption of 10 CFR 50.69 (EPID No. L-2024-LLA-0025)

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12/30/2024 3:23:12 PM Received Date:

12/30/2024 3:23:16 PM From:

Scott Wall Created By:

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"Helen L Levendosky" <hllevendosky@aep.com>

Tracking Status: None "Joe Tanko" <jmtanko@aep.com>

Tracking Status: None "Michael K. Scarpello" <mkscarpello@aep.com>

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