ML22075A339
| ML22075A339 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/15/2022 |
| From: | Mahesh Chawla NRC/NRR/DORL/LPL4 |
| To: | Ronnie Garcia Energy Northwest |
| References | |
| L-2021-LLA-0207 | |
| Download: ML22075A339 (8) | |
Text
From: Chawla, Mahesh Sent: Tuesday, March 15, 2022 1:03 PM To: Garcia, Richard M.
Cc: Hughey, John; Iqbal, Naeem; Hyslop, JS; Zhao, Jack; Hernandez, Raul; Brown, Adrienne; Hsu, Kaihwa; Huang, Jason; Ma, John; Bhatt, Santosh; Sun, Summer; Young, Austin; Quinlan, Kevin; Neuhausen, Alissa; Tetter, Keith; Goel, Vijay K; Widrevitz, Dan; Dinh, Thinh
Subject:
Columbia Generating Station - Audit Questions - Regulatory Audit of License Amendment Request to Adopt 50.69 - Categorization Process - EPID L-2021-LLA-0207 Attachments: CSG 5069 Audit Questions.docx
Dear Mr. Garcia,
By letter dated November 9, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21314A224), Energy Northwest (the licensee) submitted a license amendment request (LAR) for the Columbia Generating Station (CGS) to adopt Title 10 of the Code of Federal Regulations (CFR) Section 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. A virtual audit has been scheduled with the representatives of Energy Northwest on March 30th and 31st, 2022, to discuss the subject application. In support of the audit discussion, the U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the LAR and developed the attached audit questions. If you have any further questions, you can contact me. Thanks
Mahesh Chawla, Project Manager Licensing Branch, LPL 4 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Phone: (301) 415-8371 Email: Mahesh.Chawla@nrc.gov
OFFICEDORL/LPL4 /PM DORL/LPL4/BC
JDixon-NAMEMChawla Herrity
DATE3/14/2022 3/15/2022
Hearing Identifier: NRR_DRMA Email Number: 1557
Mail Envelope Properties (SA1PR09MB841534DEEA8DC51B7D9C874BF1109)
Subject:
Columbia Generating Station - Audit Questions - Regulatory Audit of License Amendment Request to Adopt 50.69 - Categorization Process - EPID L-2021-LLA-0207 Sent Date: 3/15/2022 1:03:01 PM Received Date: 3/15/2022 1:03:01 PM From: Chawla, Mahesh
Created By: Mahesh.Chawla@nrc.gov
Recipients:
"Hughey, John" <John.Hughey@nrc.gov>
Tracking Status: None "Iqbal, Naeem" <Naeem.Iqbal@nrc.gov>
Tracking Status: None "Hyslop, JS" <JS.Hyslop@nrc.gov>
Tracking Status: None "Zhao, Jack" <Jack.Zhao@nrc.gov>
Tracking Status: None "Hernandez, Raul" <Raul.Hernandez@nrc.gov>
Tracking Status: None "Brown, Adrienne" <Adrienne.Driver@nrc.gov>
Tracking Status: None "Hsu, Kaihwa" <Kaihwa.Hsu@nrc.gov>
Tracking Status: None "Huang, Jason" <Jason.Huang@nrc.gov>
Tracking Status: None "Ma, John" <John.Ma@nrc.gov>
Tracking Status: None "Bhatt, Santosh" <Santosh.Bhatt@nrc.gov>
Tracking Status: None "Sun, Summer" <Summer.Sun@nrc.gov>
Tracking Status: None "Young, Austin" <Austin.Young@nrc.gov>
Tracking Status: None "Quinlan, Kevin" <Kevin.Quinlan@nrc.gov>
Tracking Status: None "Neuhausen, Alissa" <Alissa.Neuhausen@nrc.gov>
Tracking Status: None "Tetter, Keith" <Keith.Tetter@nrc.gov>
Tracking Status: None "Goel, Vijay K" <Vijay.Goel@nrc.gov>
Tracking Status: None "Widrevitz, Dan" <Dan.Widrevitz@nrc.gov>
Tracking Status: None "Dinh, Thinh" <Thinh.Dinh@nrc.gov>
Tracking Status: None "Garcia, Richard M." <rmgarcia@energy-northwest.com>
Tracking Status: None
Post Office: SA1PR09MB8415.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1153 3/15/2022 1:03:01 PM CSG 5069 Audit Questions.docx 39127
Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
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,
ENERGY NORTHWEST
COLUMBIA GENERATING STATION
DOCKET NO. 50-397
By letter dated November 9, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21314A224), Energy Northwest (the licensee) submitted a license amendment request (LAR) for the Columbia Generating Station (CGS) to adopt Title 10 of the Code of Federal Regulations (CFR) Section 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The U.S.
Nuclear Regulatory Commission (NRC) staff has reviewed the LAR and request additional information in order to complete the review.
APLA Audit Question 01 - PRA Model Update Process
10 CFR 50.69(e) requires licensees to update the probabilistic risk assessment (PRA) model used for categorization when changes to the as-built, as-operated plant occur.
Section 3.2.7 of the LAR states that an unscheduled update will be performed when significant impacts to the 10 CFR 50.69 basic event importance measures occur. It is unclear to the NRC staff what constitutes a significant impact with regard to importance measures. Explain what constitutes a significant impact that would initiate a PRA model update and provide justification that this threshold does not adversely impact the categorization process.
APLA Audit Question 02 - Total Risk and Accounting for the SOKC
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, provides the risk acceptance guidelines for total core damage frequency (CDF) (1E-04 per year) and large early release frequency (LERF) (1E-05 per year). NRC staff notes based on RG 1.174 and Section 6.4 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Infomed Decisionmaking, for a Capability Category II risk evaluation, 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 that are explicitly input to the PRA models, including explicit consideration of the state-of-knowledge correlation (SOKC) between events. In general, the point estimate CDF and LERF obtained by quantification of the cutset probabilities using mean values for each basic event probability do not produce a true mean of the CDF and LERF.
Under certain circumstances, a formal knowledge SOKC is not important (i.e., the risk results are well below the acceptance guidelines).
to the LAR provides the CDF and LERF values for CGS that appear to be point estimates which are likely lower than the mean CDF and LERF values. The total LERF value of 9.1E-05 per year does approach the RG 1.174 guidelines of 1E -05 per year for LERF.
Given this information, demonstrate that the total mean internal events (including internal flooding), fire, and seismic LERF is in conformance with RG 1.174 risk acceptance guidelines (i.e., LERF < 1E-05 per year).
APLB Audit Topic
Paragraphs 50.69(c)(1)(i) and (ii) of 10 CFR requires a licensees PRA be of sufficient quality and level of detail to support the structures, systems, and components (SSC) categorization process, and requires 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, Rev.0, SSC Categorization Guideline, 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 the SSC(s) importance.
Section 3.2.8 of the LAR describes the licensees uncertainty determination process for this application. Specifically, 42 items were identified for consideration and screened using four criteria. After this screening, only three items were determined to be key sources of uncertainty to be addressed by sensitivity studies during the categorization process.
Therefore, it is recommended to have staff better understand the process, staff requests a presentation of this process with examples at each stage. The licensees notebook, ENGNW-00554-REPT-001, Revision 0 does provide some insights, but the staff would like to gain a better understanding.
APLC Audit Question 01 - Dispositions of PRA Model Assumptions and Sources of Uncertainty Paragraphs (c)(1)(i) and (ii) of 10 CFR 50.69 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 applicable sensitivity studies identified in the characterization of PRA adequacy should be considered.
The NRC staff notes that the uncertainty assessment may identify the need for additional sensitivity studies beyond those presented in Table 5-2 and Table 5-3 of NEI 00-04. LAR identifies the key assumptions and sources of uncertainty for the PRA and provides dispositions for each key assumption/source of uncertainty. The NRC staff reviewed CGS Impact of Model Uncertainty to 50.69 Categorization Process, [ENGNW-00554-REPT-001, Revision 0], which does not appear to adequately address certain sources of uncertainty that have the potential to impact 10 CR 50.69 categorization.
Regarding the seismic early release timing uncertainty in Table 3 of ENGNW-00554-REPT-001, a sensitivity study determined that when large intermediate releases are included to address delayed evacuation, an increase in LERF risk from 5.2E-06 per year to 1.4E-05 per year is estimated. However, the report evaluation states that LERF modeling is conservative since 43 percent of LERF releases are unscrubbed and prompt. It is unclear to the NRC staff how any potential conservatism in the unscrubbed prompt releases offsets the uncertainty in timing of evacuation following a seismic event. Therefore, address the following:
a) Discuss how the LERF modeling conservatisms address the uncertainty in timing of evacuation following a seismic event. Include in this discussion the specific conservatisms and their impact on offsetting the uncertainty in timing of evacuation following a seismic event.
b) Provide justification that uncertainty in timing of evacuation following a seismic event is not a key source of uncertainty for this application (i.e., does not significantly impact SSC categorization).
c) As an alternative to Part (b), propose a mechanism, preferably consistent with NEI 00-04, to ensure the uncertainty of late evacuations is addressed in the CGS categorization program.
APLC Audit Question 02 - Other External Hazard Screening Criteria for Sand and Dust Storms NEI 00-04, Rev. 0, Section 5, Component Safety Significance Assessment states, If the plant does not have an external hazards PRA, then it is likely to have an external hazards screening evaluation that was performed to support the requirements of the IPEEE. NEI 00-04, Rev. 0, also states in Section 3.3.2, Other Risk Information (including other PRAs and screening methods), that the characterization of the adequacy of risk information should include a basis for why the other risk information adequately reflects the as-built, as-operated plant.
Attachments 4 and 5 of the LAR provide the screening results of other external hazards for 10 CFR 50.69 categorization and Section 3.2.4 of the LAR enclosure, Evaluation of the Proposed Change, concludes that all external hazards, except for internal flooding, internal fire, and seismic activity, were screened from applicability to CGS. Attachment 4 of the LAR indicates that criterion C1 (event damage potential is less than events for which plant is designed) and criterion C4 (event is included in the definition of another event) were used to screen the sand or dust storm hazard. The LAR states that Sand or dust storm is bounded by a postulated volcanic ash event. In Attachment 4 of the LAR, the screening justification for volcanic activity states Enhancements to seismic monitoring capability over the past decade, as well as ash fall monitoring by the Volcanic Ash Advisory Center over the past several decades, improved the timeliness and detail of information available to CGS in advance of potential ash fall. Therefore, there is adequate time to take mitigating actions (e.g., plant power reductions or shutdown). It is unclear to the NRC staff that the sand or dust storm event is bounded by a postulated volcanic ash event since the plant mitigation capability described for volcanic activity may not be applicable to a sand or dust storm, and therefore, that the assumptions that resulted in the screening are appropriate for screening this hazard from 10 CFR 50.69 categorization.
a) Provide justification for stating that the sand or dust storm hazard is bounded by a postulated volcanic ash event. The justification should include an explanation for the basis for determining that the screening analysis adequately reflects the as-built, as-operated plant.
b) Discuss assumptions involved in the screening of the sand and dust storm hazard, including any SSCs credited for the screening and explain how it will be ensured that these assumptions continue to remain valid during the implementation of 10 CFR 50.69.