ML21258A038

NRR E-mail Capture - Final - Request for Additional Information - Callaway, Unit 1 - LAR to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors- EPID L-2020-LLA-023

ML21258A038
Person / Time
Site:	Callaway
Issue date:	09/14/2021
From:	Mahesh Chawla NRC/NRR/DORL/LPL4
To:	Elwood T Ameren Missouri, Ameren Union Electric Co
References
L-2020-LLA-0235
Download: ML21258A038 (8)

Text

From: Chawla, Mahesh Sent: Tuesday, September 14, 2021 5:11 PM To: 'Elwood, Thomas B' Cc: Dixon-Herrity, Jennifer; Havertape, Joshua; Circle, Jeff; Pascarelli, Robert; Wu, De; Grenier, Bernard; Tetter, Keith; Rosenberg, Stacey; Quinlan, Kevin; Heeszel, David; Hayes, Barbara; Lee, Brian; Nold, David; Robinson, Jay; Jenkins, Joel; Ashcraft, Joseph; Hsu, Kaihwa; Farnan, Michael; Pettis, bob; Cumblidge, Stephen; Goel, Vijay

Subject:

FINAL - Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors"- EPID L-2020-LLA-0235 Attachments: Callaway 50.69 RAIs.docx

Dear Mr. Elwood,

By application dated October 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20304A454), Union Electric Company, dba Ameren Missouri (the licensee), submitted a license amendment request for Callaway Plant, Unit No. 1 (Callaway). The proposed amendment would modify the Callaway licensing basis by the addition of a license condition (i.e., License Condition 2.C.(19)), to allow for the implementation of the provisions of 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. On February 18, 2021 (ADAMS Accession No. ML21039A222), the U.S.

Nuclear Regulatory Commission (NRC) staff issued an audit plan that conveyed the intent to conduct a regulatory audit to support its review of the subject license amendment. The NRC staff and the representatives of Ameren Missouri, participated in the regulatory audit, conducted from May 11, 2021 through May 13, 2021. The NRC issued an audit summary in a letter dated June 9, 2021 (ADAMS Accession No. ML21139A022). The licensee provided additional/revised information in a supplement dated July 29, 2021 (ADAMS Accession No. ML21210A025). The NRC staff has reviewed the supplemental information provided after completion of the regulatory audit, and has determined that the attached additional information is required in order to complete the review of the subject LAR.

The attached RAIs were sent to you via email on September 2, 2021 as the draft RAIs and requested you to set up a clarification call with the NRC staff. On September 14, 2021 you informed NRC that your staff understands the requested information and has determined that there is no need for a clarification call with the NRC staff. You have also agreed to provide the supplemental response to the requested information within 30 calendar days of the receipt of this email. Thanks Sincerely, Mahesh Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ph: 301-415-8371

OFFICE DORL/LPL4/PM DORL/LPL4/BC JDixon-NAME MChawla Herrity DATE 9/14/2021 9/14/2021

Hearing Identifier: NRR_DRMA Email Number: 1347 Mail Envelope Properties (SA1PR09MB8415A687B9B1E78AA66851B1F1DA9)

Subject:

FINAL - Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors"- EPID L-2020-LLA-0235 Sent Date: 9/14/2021 5:11:18 PM Received Date: 9/14/2021 5:11:19 PM From: Chawla, Mahesh Created By: Mahesh.Chawla@nrc.gov Recipients:

"Dixon-Herrity, Jennifer" <Jennifer.Dixon-Herrity@nrc.gov>

Tracking Status: None "Havertape, Joshua" <Joshua.Havertape@nrc.gov>

Tracking Status: None "Circle, Jeff" <Jeff.Circle@nrc.gov>

Tracking Status: None "Pascarelli, Robert" <Robert.Pascarelli@nrc.gov>

Tracking Status: None "Wu, De" <De.Wu@nrc.gov>

Tracking Status: None "Grenier, Bernard" <Bernard.Grenier@nrc.gov>

Tracking Status: None "Tetter, Keith" <Keith.Tetter@nrc.gov>

Tracking Status: None "Rosenberg, Stacey" <Stacey.Rosenberg@nrc.gov>

Tracking Status: None "Quinlan, Kevin" <Kevin.Quinlan@nrc.gov>

Tracking Status: None "Heeszel, David" <David.Heeszel@nrc.gov>

Tracking Status: None "Hayes, Barbara" <Barbara.Hayes@nrc.gov>

Tracking Status: None "Lee, Brian" <Brian.Lee@nrc.gov>

Tracking Status: None "Nold, David" <David.Nold@nrc.gov>

Tracking Status: None "Robinson, Jay" <Jay.Robinson@nrc.gov>

Tracking Status: None "Jenkins, Joel" <Joel.Jenkins@nrc.gov>

Tracking Status: None "Ashcraft, Joseph" <Joseph.Ashcraft@nrc.gov>

Tracking Status: None "Hsu, Kaihwa" <Kaihwa.Hsu@nrc.gov>

Tracking Status: None "Farnan, Michael" <Michael.Farnan@nrc.gov>

Tracking Status: None "Pettis, bob" <Robert.Pettis@nrc.gov>

Tracking Status: None "Cumblidge, Stephen" <Stephen.Cumblidge@nrc.gov>

Tracking Status: None "Goel, Vijay" <Vijay.Goel@nrc.gov>

Tracking Status: None

"'Elwood, Thomas B'" <TElwood@ameren.com>

Tracking Status: None Post Office: SA1PR09MB8415.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 2320 9/14/2021 5:11:19 PM Callaway 50.69 RAIs.docx 31721 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

CALLAWAY PLANT, UNIT NO. 1 LICENSE AMENDMENT TO ADOPT 10 CFR 50.69 RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS REQUEST FOR ADDITIONAL INFORMATION By application dated October 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20304A454), Union Electric Company, dba Ameren Missouri (the licensee), submitted a license amendment request (LAR) for Callaway Plant, Unit No. 1 (Callaway). The proposed amendment would modify the Callaway licensing basis by the addition of a license condition (i.e., License Condition 2.C.(19)), to allow for the implementation of the provisions of 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. On February 18, 2021 (ADAMS Accession No. ML21039A222), the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan that conveyed the intent to conduct a regulatory audit to support its review of the subject license amendment. The NRC staff and the representatives of Ameren Missouri, participated in the regulatory audit, conducted from May 11, 2021 through May 13, 2021. The NRC issued an audit summary in a letter dated June 9, 2021 (ADAMS Accession No. ML21139A022). The licensee provided additional/revised information in a supplement dated July 29, 2021 (ADAMS Accession No. ML21210A025). The NRC staff has reviewed the supplemental information provided after completion of the regulatory audit, and has determined that additional information is required in order to complete the review of the subject LAR. Please arrange a clarification teleconference with the NRC staff at your earliest possible.

RAI 01 - Use of Mean Core Damage Frequency and Large Early Release Frequency Values and Consideration of the State of Knowledge Correlation Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ADAMS Accession No ML17317A256) and Section 6.4 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decisionmaking, dated March 2017 (ADAMS Accession No. ML17062A466), for a Capability Category II risk evaluation, indicate that the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to are the means of the probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly reflected in the PRA models. In general, the point estimate core damage frequency (CDF) and large early release frequency (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 and LERF. 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 (i.e., the risk results are well below the acceptance guidelines). of the updated LAR, in response to Audit Question APLA/APLC 05 Part (c) provides updated quantification results based on PRA update 9.01 using mean values which account for the SOKC showing the total CDF and LERF meet the RG 1.174 risk acceptance guidelines. However, the updated LAR did not state whether fire, seismic and wind PRA Enclosure

parameters that should be correlated to account for SOKC were included such as circuit failure probabilities, suppression probabilities, and ignition frequencies for the fire PRA. The response also states that the mean values are conservative because of the inability to fully post-process with ACUBE [advanced cutset upper bound estimator]. It is not clear how this conservatism (i.e., that the mean values are 19 and 27 percent higher than the point estimate values for CDF and LERF respectively) will be addressed for the 10 CFR 50.69 program, given that it would appear to impact to 10 CFR 50.69 risk-informed categorization. Therefore, address the following:

a) Identify the parameters derived from the same data (other than for same type code data) that were correlated for the fire, seismic, and high wind PRAs, and justify that these parameters are sufficient to estimate the SOKC.

b) Discuss how the SOKC will be treated for the 10 CFR 50.69 program consistent with NUREG-1855, Revision 1, when the risk increase associated with SOKC is considered.

RAI 02 - Treatment of Sensitive Electronics in the Fire PRA RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, states NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. The relatively extensive and detailed reviews of fire PRAs undertaken in support of LARs to transition to National Fire Protection Association (NFPA)-805, Performance-Based Standard for Light Water Reactor Electric Generating Plants, determined that implementation of some of the complex fire PRA methods often used non-conservative and over-simplified assumptions to apply the method to specific plant configurations. Some of these issues were not always identified in facts and observations (F&Os) by the peer review teams but are considered potential key assumptions by the NRC staff because using more defensible and less simplified assumptions could substantively affect the fire risk and fire risk profile of the plant.

LAR Section 3.2.2 states that the numerous new or revised fire PRA guidance documents issued since Callaway was approved to implement NFPA-805 are being addressed through the PRA maintenance and update process. With regards to use of Frequently Asked Question (FAQ)13-0004, Clarifications on Treatment of Sensitive Electronics (ADAMS Accession No. ML13322A085), Attachment 8 of the updated LAR states that [w]ithout explicitly citing FAQ 13-0004 for the treatment of sensitive electronics, the fire PRA does implement the salient conclusion that a generic screening heat flux damage threshold for thermoset cables, as observed on the outer surface of the cabinet, can be used as a conservative surrogate for assessing the potential for thermal damage to solid-state and sensitive electronics within an electrical panel (cabinet). The LAR, however, does not address the two caveats cited in FAQ 13-0004 that can invalidate this approach which are (1) sensitive electronics mounted on the surface of the cabinet where it can be exposed to the convective or radiant energy of a fire, and (2) the presence of a louver or other typical ventilation means. It appears to the NRC staff that this source fire PRA modeling uncertainty could have the potential to impact the application.

Therefore, address the following:

a) Explain how sensitive electronics are modeled in the fire PRA for sensitive electronics cabinet configurations (i.e., wall mounted electronics or ventilation) that invalidate the FAQ 13-0004 approach described in the LAR (i.e., using the heat flux damage threshold for thermoset cables for sensitive electronics inside a cabinet). Include confirmation that the approach is consistent with the guidance in FAQ 13-0004.

b) If the approach cannot be confirmed to be consistent with the guidance in FAQ 13-0004, then, justify that it has an inconsequential impact on 10 CFR 50.59 risk-informed categorization. Alternatively, explain how the uncertainty associated with this modelling treatment will be addressed in the 10 CFR 50.69 program.

RAI 03 - Seismic PRA Modeling RG 1.174, Revision 3, states, in part, that the plant-specific PRA supporting the licensees proposals has been demonstrated to be acceptable.

Review of PRA-SEISMIC-QUANT (described in Audit Question APLA/APLC-03 Response in , Attachment 8 to the LAR supplement dated July 29, 2021), Table 4-2 indicates that the highest ACUBE conditional core damage probability (CCDP) is 0.9 and its note (2) states that the value in the column includes the plant availability factor (PAF) of 0.9. However, Table 4-3 shows that ACUBE conditional large early release probability (CLERP) is close to 1.

It is not clear why the value in the column does not include the PAF of 0.9, since they are very similarly calculated as a ratio of either ACUBE CDF or ACUBE LERF to seismic frequency. In addition, on page 54 it says that both CCDP and CLERP approach the PAF of 0.9 for the last hazard interval bin. However, Figures 4-3, 4-4, and 4-5 do not demonstrate this. Instead, Figures 4-3 and 4-5 show CCDP more than 0.9, while Figures 4-4 and 4-5 show CLERP is close to 1.

The last bin for seismic CDF represents 0.88 g peak ground acceleration, covering a seismic hazard from 0.8 to 10 g. Provide a sensitivity study to demonstrate no impact on the final seismic CDF when this bin is divided into several sub-bins (for example, 0.8-1.0 g, 1.0-1.2 g, 1.2-1.4 g, 1.4-1.6 g, 1.6-2.0 g and 2.0-10 g).

RAI 04 - Seismic Fragility Analysis Uncertainty Section 5 of Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC [Structure, System, and Component] Categorization Guideline, provides guidance for performing sensitivity studies for each PRA model to address the uncertainty associated with those models. Specifically, Sections 5.1, 5.2, and 5.3 provide guidance for such sensitivities for the internal events, fire and seismic PRA, respectively. 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.

The high seismic CDF and LERF values relative to the CDF and LERF values from the other hazards including internal events suggest that the uncertainty in seismic PRA modeling involving the level of detail used to model fragility could potentially impact 10 CFR 50.69 risk-informed categorization. It is not clear from the discussion provided in Attachment 8 of the updated LAR in response to Audit Question APLC 04 on the level of fragility analyses performed for four dominant CDF and LERF importance contributors (i.e., seismically induced loss of offsite power, failure of service water, failure of steam generator supports and soil failure) whether the Callaway seismic CDF and LERF values could be further reduced by further refining the fragility analyses for other SSCs. NRC staff observes that since the point estimate seismic CDF of 5.59E-05 per year presented in Section 6 of the Callaway seismic PRA report dated July 10, 2020 (ADAMS Accession No. ML20192A244) the calculated point estimate seismic CDF value has been reduced to 4.01E-05 based on refinements according to of the updated 10 CFR 50.69 LAR. An overly conservative seismic PRA model

could skew the integrated importance measures calculated for 10 CFR 50.69 categorization.

Importance measures provide a relative measure of risk importance and, therefore, if the importance of certain SSCs is significantly overestimated, then the importance of other SSCs will be underestimated. In light of these observations, address the following:

a) Justify that not using more refined fragility analysis for certain important SSCs will not have a consequential impact on 10 CFR 50.69 risk-informed categorization. Include in the discussion how the SSC fragility analysis for important SSCs beyond just the top four contributors is performed.

b) If it cannot be justified that using more refined fragility analysis for certain important SSCs will not have a consequential impact on 10 CFR 50.69 categorization, then explain how this modeling uncertainty will be addressed for the 10 CFR 50.69 program.