ML22308A096
| ML22308A096 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/15/2022 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Schuetz R Energy Northwest |
| Chawla M | |
| References | |
| EPID L-2021-LLA-0207 | |
| Download: ML22308A096 (18) | |
Text
December 15, 2022 Mr. Robert Schuetz Chief Executive Officer Energy Northwest 76 North Power Plant Loop P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352
SUBJECT:
COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT NO. 269 RE: LICENSE AMENDMENT TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2021-LLA-0207)
Dear Mr. Schuetz:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 269 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Technical Specifications (TS) in response to your application dated November 9, 2021, as supplemented by letters dated May 10, 2022, and October 17, 2022.
The amendment revises the Columbia Renewed Facility Operating License No. NPF-21 to add a new license condition to allow for the implementation of Title 10 of the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation).
R. Schuetz A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 269 to NPF-21
- 2. Safety Evaluation cc: Listserv
ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 License No. NPF-21
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Energy Northwest (the licensee), dated November 9, 2021, as supplemented by letters dated May 10, 2022, and October 17, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. NPF-21 is hereby amended to add paragraph 2.C.(37) to read as follows:
(37) 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Energy Northwest 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 (SSC) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 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 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. 269 dated December 15, 2022. Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-21 Date of Issuance: December 15, 2022 Jennifer L.
Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.12.15 13:58:39 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 269 COLUMBIA GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of Renewed Facility Operating License No. NPF-21 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT
Renewed License No. NPF-21 Amendment No. 225, 266 269 (34)
Deleted (35)
The licensee's FSAR, as updated with the license renewal FSAR supplement submitted pursuant to 10 CFR 54.21(d) and supplemented with Appendix A of NUREG-2123 with the exception of Commitments Nos. 55, 56, 57, and 71, and as revised pursuant to the criteria set forth in 10 CFR 50.59, describes certain future programs and activities to be completed before the period of extended operation. Energy Northwest shall complete these activities no later than July 20, 2023 and shall notify the NRC in writing when implementation of these activities is complete.
(36)
To prevent lateral motion of the core plate, the licensee shall install core plate wedges around the periphery of the core plate within the shroud on or before December 20, 2021. Upon completion of the core plate wedge installation, the licensee shall submit a written report to the NRC staff summarizing the results of the installation. The licensee shall also submit a written report regarding any corrective action taken related to core plate rim hold-down bolts or core plate wedges and the results of extent of condition reviews on or before December 20, 2021.
(37) 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Energy Northwest 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 (SSC) using:
Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 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 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. 269 dated December 15, 2022.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By letter dated November 9, 2021 (Reference [1]), as supplemented by letter dated May 10, 2022 (Reference [2]), and October 17, 2022 (Reference [3]), Energy Northwest (the licensee) submitted a license amendment request (LAR) for the Columbia Generating Station (CGS). The licensee proposed a license condition to the Renewed Facility Operating License (RFOL) to allow the implementation 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.
The license condition will allow for adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation).
The supplemental letters dated May 10, 2022, and October 17, 2022, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on February 22, 2022 (87 FR 9651).
2.0 REGULATORY EVALUATION
2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design basis functions. For SSCs categorized as low safety significance (LSS),
alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.
10 CFR 50.69 contains requirements regarding how a licensee categorizes SSCs using a risk-informed process; adjusts treatment requirements consistent with the relative significance of the SSC; and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four RISC categories:
RISC-1: Safety-related SSCs that perform safety significant functions RISC-2: Nonsafety-related SSCs that perform safety significant functions RISC-3: Safety-related SSCs that perform low safety significant functions RISC-4: Nonsafety-related SSCs that perform low safety significant functions SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.
Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or potentially enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on SSCs categorized as HSS.
2.2 Regulatory Guides The NRC staff considered the following regulatory guidance during its review of the proposed changes:
RG 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance" (Reference [4])
RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference [5])
RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference [6])
RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference [7])
NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference [8]), as endorsed by RG 1.201 for trial use with clarifications and describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.
Sections 2 through 10 of NEI 00-04 describe the following elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:
Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).
Sections 3, 4, 5, and 7 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).
Section 6 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).
Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).
Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).
Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).
Additionally, Section 11 of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f). Section 12 of NEI 00-04 provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e). Maintaining change control and periodic review provides confidence that all aspects of the program reasonably reflect the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii).
3.0 TECHNICAL EVALUATION
3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions on the use of PRA findings and risk insights in support of changes to a plants licensing basis, is to show that the proposed licensing basis changes meet the five key principles stated in section C of RG 1.174, Revision 3.
3.2 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles of RG 1.174, Revision 3.
3.2.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations The SSCs are classified as having either HSS functions (i.e., RISC-1 and RISC-2 categories) or LSS functions (i.e., RISC-3 and RISC-4 categories). For HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements from these SSCs). For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For RISC-3 SSCs, licensees can replace special treatment with an alternative treatment. For RISC-4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.
The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1, and the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process. The NRC staffs review, as documented in this safety evaluation (SE), used the framework provided in RG 1.174, Revision 3, and NEI 00-04, Revision 0.
In Section 3.1 of the LAR, the licensee stated that it will implement the risk-informed categorization process in accordance with NEI 00-04, Revision 0. The licensee provided further
discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the NEI 00-04, Revision 0.
The regulatory requirements in 10 CFR 50.69, the monitoring outlined in NEI 00-04, Revision 0, and clarifications in RG 1.201, Revision 1 ensure that the SSC categorization process is sufficient to assure that the SSC functions continue to be met and that any performance deficiencies will be identified, and appropriate corrective actions taken. The licensees SSC categorization program includes the appropriate elements prescribed in NEI 00-04, Revision 0 to assure that SSCs are appropriately categorized consistent with 10 CFR 50.69. The staff performed a detailed review of specific elements of the licensees SSC categorization process where necessary to confirm consistency with the NEI 00-04 guidance. In light of the above, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision-making prescribed in RG 1.174, Revision 3.
3.2.2 Key Principle 2: Licensing Basis Change is Consistent with the DID Philosophy The NRC staff considered the following from RG 1.174, Revision 3, in evaluating how the LB change is maintained for the DID philosophy:
Preserve a reasonable balance among the layers of defense.
Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.
Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.
Preserve adequate defense against potential common-cause failures.
Maintain multiple fission product barriers.
Preserve sufficient defense against human errors.
Continue to meet the intent of the plants design criteria.
In Section 3.1.1 of the LAR, the licensee stated that it would require an SSC to be initially categorized as HSS based on the DID assessment performed in accordance with NEI 00-04, Revision 0. Based on the above, the staff concludes that the proposed change is consistent with the DID philosophy described in key principle 2 of RG 1.174, Revision 3, and is, therefore, acceptable. The NRC staff finds that the licensee's process is consistent with the NRC-endorsed guidance in NEI 00-04 and would meet the 10 CFR 50.69(c)(1)(iii) criterion that requires DID to be maintained.
3.2.3 Key Principle 3: LB Change Maintains Sufficient Safety Margins The regulations in 10CFR50.69(c)(1)(iv) require in part that an evaluation be performed to provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of the SSC to assure that sufficient safety margins are maintained. With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the licensing basis (e.g., Updated Final Safety Analysis Report, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data. RG 1.174, Revision 3 provides guidelines for making an assessment including evaluations to ensure the categorization of the SSC does not adversely
affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.
The staff evaluated the categorization process described in Section 3.1.1 of the LAR and found that it contains the elements necessary to assure that SSC design basis function(s) as described in the plants LB, including the Updated Final Safety Analysis Report and Technical Specifications Bases, does not change and continues to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the staff concludes that the licensee has established a program to ensure sufficient safety margins are maintained in accordance with the third key principle of RG 1.174, Revision 3 and would therefore meet 10 CFR 50.69(c)(1)(iv).
3.3 Risk-Informed Assessment 3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04, Revision 0, endorsed by RG 1.201, Revision 1 address the fourth and fifth key principles in risk-informed decision-making, pertaining to the assessment of the change in risk and monitoring the impact of the LB change.
A summary of how the licensees SSC categorization process is consistent with the guidance and methodology prescribed in NEI 00-04, Revision 0, and RG 1.201, Revision 1 is provided in the sections below.
In Sections 3.2.1, 3.2.2, and 3.2.3 of the LAR, the licensee described that the CGS categorization process uses PRA modeled hazards to assess risks for the internal events (includes internal floods), internal fires, and seismic events. For the other risk contributors, the licensee's process uses the following non-PRA methods to characterize the risk:
Other External Hazards: Screening analysis performed for Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)
(Generic Letter No. 88-20, Supplement 4) (Reference [9]) updated using criteria from Part 6 of the American Society of Mechnical Engineers (ASME) and American Nuclear Society (ANS) 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, (the PRA Standard) as endorsed by the NRC (Reference [10]).
Shutdown Events: Safe Shutdown Risk Management program consistent with Nuclear Utility Management and Resources Council (NUMARC) 91-06, "Guidelines for Industry Actions to Assess Shutdown Management" (Reference [11]).
Passive Components: ANO-2 passive categorization methodology, Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, (Reference [12]).
The approaches and methods proposed by the licensee to address internal events, internal fires, seismic events, other external hazards, DID, and shutdown events are consistent with the approaches and methods included in the guidance in NEI 00-04, Revision 0. The licensee proposed to use the shutdown safety assessment process based on NUMARC 91-06. The shutdown safety assessment method is consistent with the guidance in NEI 00-04. The non-
PRA method for the categorization for passive components is consistent with the ANO-2 passive categorization methodology.
3.3.1.1 Scope of the PRA The CGS PRA is comprised of a full-power, Level 1, internal events PRA (IEPRA), fire PRA (FPRA), and seismic PRA (SPRA), which evaluates the core damage frequency (CDF) and large early release frequency (LERF) risk metrics. The licensee discussed in Section 3.2 of the LAR, that the IEPRA and SPRA models have been assessed against RG 1.200, Revision 2, and that the FPRA has been assessed against RG 1.200, Revision 3.
The NRC staff reviewed two aspects of the PRA with regard to the impact of the proposed changes on plant operational risk: (1)scope and acceptability of the PRA models and their application to the proposed changes, and (2)a review of the PRA results and insights described in the licensees application.
Evaluation of Modeled PRAs In Section 3.2 of the Enclosure of the LAR, the licensee confirmed that the IEPRA and SPRA models had been peer reviewed in accordance with RG 1.200, Revision 2, and the FPRA in accordance with RG 1.200, Revision 3. The licensee stated that closure of the Facts and Observations (F&Os) resulting from the peer reviews was performed using an independent assessment process. The NRC staff confirmed that the licensee performed closure of the F&Os consistent with Final Revision of Appendix X to NEI 05-04/07-12/12-06: 'Close-Out of Facts and Observations (F&Os), February 21, 2017 (Reference [13]), as endorsed in RG 1.200, Revision 3. In Attachment 6 of the LAR, the licensee provided a brief discussion and list of the key assumptions and sources of uncertainty, along with treatment or disposition for the application of 10 CFR 50.69. The NRC staff concludes the licensees modeled PRAs have been peer reviewed using endorsed guidance, F&Os closed, and the key assumptions and sources of uncertainty have been identified and dispositioned appropriately.
The licensee states in Section 3.2.9 of the LAR that the CGS PRA model credits diverse and flexible coping strategies (FLEX) equipment or FLEX strategies in the PRA. To address concerns with the uncertainty of the methods used to model FLEX equipment and operator actions described in the NRC staff memorandum of May 30, 2017, "Assessment of the Nuclear Energy Institute 16-06, 'Crediting Mitigating Strategies in Risk-Informed Decision Making,'
Guidance for Risk-Informed Changes to Plants Licensing Basis" (Reference [14]), the licensee provided the results of a sensitivity study in Section 3.2.9 of the LAR that demonstrated that the modeling of FLEX equipment in all three PRA models had a negligible impact on CDF and LERF and does not constitute a key source of uncertainty for this application.
The NRC staff concluded that the licensees credit for FLEX equipment in the LAR is appropriate because the licensee used consensus human reliability analysis methodologies and practices and acceptable failure rates and performed sensitivity studies to assess the impact on the LAR.
The NRC staff reviewed the PRA models peer-review history provided by the licensee in the Enclosure of the LAR, as supplemented. The NRC staff further considered the key assumptions and sources of uncertainty identified by the licensee and the licensees crediting of FLEX. The NRC staff finds the CGS scope and acceptability of the modeled IEPRA, FPRA, and SPRA to
be commensurate with their use in the LAR, for use in the integrated decision-making process, and consistent with RG1.174.
3.3.1.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization As part of its proposed integrated decision-making process to categorize SSCs according to safety significance, the licensee has proposed to use a non-PRA method to consider other external hazards. 10 CFR 50.69(c)(1)(ii) and 50.69(b)(2)(ii) permit the use of non-PRA methods in a risk-informed categorization process.
Other External Hazards This hazard category includes all non-seismic external hazards such as high winds, external floods, transportation, nearby facility accidents, and other hazards.
The SE report for the CGS "Review of Columbia Generating Station Individual Plant Examination of External Events Submittal (Tac No. M83695)," February 26, 2001, (Reference
[15]) concludes that high winds, floods, transportation, and other external events areas were eliminated based on either compliance with NUREG-75/08, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants, LWR [Light Water Reactor] edition, December 1975 (Reference [16]) criteria or on the basis of a bounding probabilistic assessment resulting in a CDF estimate less than 10-6 per reactor year, i.e., below the screening criteria in NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (Reference [17]).
In Attachment 4 of the LAR the licensee provided the plant-specific external hazards screening results using the IPEEE screening process updated using endorsed criteria in the PRA Standard and current plant hazard information. In its May 10, 2022, supplement the licensee clarified the station procedures to respond to sand or dust storm conditions since the sand or dust storm event is bounded by the volcanic ash event as described in Section 5.5.2 of the CGS IPEEE.
In Section 3.2.4 of the LAR, the licensee stated, in part, that all other external hazards were screened from applicability to CGS per a plant-specific evaluation in accordance with Generic Letter 88-20 and updated to use the criteria in the PRA Standard. The licensee stated that SSCs will be evaluated during categorization of the SSC using guidance in NEI 00-04, Figure 5-6 to identify any SSCs that will be considered HSS. The NRC staff finds that CGS will assess the risk from all other external hazards consistent with NEI 00-04 as endorsed in RG 1.201, Revision 1.
In summary, the use of the CGS IPEEE results described by the licensee in the LAR, supplemental information provided, and the licensee's updated assessment of the other external hazards is consistent with Section 5 of NEI 00-04, Revision 0 as endorsed in RG 1.201, Revision 1. The NRC staff concludes that the licensee's treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).
Component Safety Significance Assessment for Passive Components In Section 3.1.2 of the LAR the licensee proposed using a categorization method, the ANO-2 passive categorization methodology, not cited in NEI 00-04, Revision 0, or RG 1.201,
Revision 1. The ANO-2 passive categorization methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference [18]). The ANO-2 passive categorization methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance is generally measured by the frequency and the consequence of, in this case, pipe ruptures. Treatment requirements (including repair/replacement) only affect the frequency of passive component failure. Categorizing solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. Therefore, the NRC staff finds that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.
In Section 3.1.2 of the LAR, the licensee stated, [t]he passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the[Alternative]
ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components.
Consistent with Alternative ANO-2-R&R-004, Class 1 pressure retaining SSCs in the scope of the system being categorized will be assigned HSS and cannot be changed by the Integrated Decision-making Panel. The NRC staff finds the licensee's proposed approach for passive categorization acceptable for the 10 CFR 50.69 SSC categorization process.
3.3.1.3 Key Principle 4 Conclusions Based on the above, the NRC staff review for the IEPRA, FPRA, and SPRA acceptability and evaluation of the use of non-PRA methods concludes that the proposed change satisfies the fourth key principle for risk-informed decision-making prescribed in RG 1.174, Revision 3.
3.3.2 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04, Revision 0 provides guidance that includes programmatic configuration control and a periodic review to ensure that all aspects of the 10 CFR 50.69 program (i.e., includes traditional engineering analyses) and PRA models used to perform the risk assessment continue to reflect the as-built, as-operated plant and that plant modifications and updates to the PRA over time are continually incorporated.
Sections 11 and 12 of NEI 00-04, Revision 0 include discussion on program documentation, change control, and periodic review. Maintaining change control and periodic review will also maintain confidence that all aspects of the 10 CFR 50.69 program and risk categorization for SSCs continually reflect the CGS as-built, as-operated plant.
The NRC staff finds the risk management process described by the licensee in the LAR is consistent with Sections 11 and 12 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and consistent with the requirements in 10 CFR 50.69(e). Based on the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision-making prescribed in RG 1.174, Revision 3.
4.0 PROPOSED CHANGE
S TO THE OPERATING LICENSE The licensee proposed the following amendment to the RFOLs for CGS. The proposed license condition states:
- 1) Energy Northwest 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 (SSC) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 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 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. 269, dated December 15, 2022.
- 2) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.
The NRC staff finds that the proposed license condition is acceptable, because: (1) it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that has previously been endorsed by the NRC; and (2) the evaluation in SE Section 3.3.1.2, finds the non-PRA methods for assessing safety significance for passive components which is not cited in NEI 00-04, to be acceptable.
The NRC staff notes that the guidance for implementing 10 CFR 50.69 provided by the Commission in the Federal Register notice dated November 22, 2004 (69 FR 68008, 68028-68029),Section III.4.10.2, Section 50.36 Technical Specifications, stated that the 10 CFR 50.69 rule does not include 10 CFR 50.36 in the list of special treatment requirements that may be replaced by the alternative 10 CFR 50.69 requirements for RISC-3 and RISC-4 SSCs when implementing a 10 CFR 50.69 license amendment. As a result, the NRC staff does not consider the TSs (including Improved Technical Specifications and the associated Technical Requirements Manual) to be part of the 10 CFR 50.69 rule. Therefore, the licensee needs to address proposed changes to its TS separately.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Washington State official was notified of the proposed issuance of the amendment on November 3, 2022. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no
significant hazards consideration published in the Federal Register on February 22, 2022 (87 FR 9651), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
[1] Dittmer, J.K., Energy Northwest letter to U.S. Nuclear Regulatory Commission, "Columbia Generating Station, Docket No. 50-397 License Amendement Request to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors,'" November 9, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21314A224).
[2] Dittmer, J.K., Energy Northwest letter to U.S. Nuclear Regulatory Commission, "Columbia Generating Station, Docket No. 50-397 Supplement to License Amendement Request to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors,'" May 10, 2022 (ML22130A591).
[3] Hauger, J.S., Energy Northwest letter to U.S. Nuclear Regulatory Commission, "Columbia Generating Station, Docket No. 50-397 Supplement to License Amendement Request to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors,'" October 17, 2022 (ML22290A251).
[4] Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance," Revision 1, May 2006 (ML061090627).
[5] Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (ML090410014).
[6] Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020 (ML20238B871).
[7] Regulatory Guide 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," January 2018 (ML17317A256).
[8] NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revsion 0, July 2005 (ML052910035).
[9] U.S. Nuclear Regulatory Commission, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), (Generic Letter 88-20, Supplement 4)," June 28, 1991, (ML031150485)NUREG-75/087, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants," LWR edition, December 1975.
[10] ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2, 2009, New York, NY (Copyright).
[11] NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management,"
December 1991 (ML14365A203).
[12] Markley, Michael, U.S. Nuclear Regulatory Commission, letter to Vice President, Operations, Arkansas Nuclear One, Entergy Operations, Inc., Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, April 22, 2009 (ML090930246).
[13]
"NEI 05-04/07-12/12-06 Appendix X: 'Close-Out of Facts and Observations (F&Os),
Washington, DC, February 21, 2017 (ML17086A431).
[14] Reisi-Fard, M., U.S. Nuclear Regulatory Commission, memorandum to Giitter, J.G., U.S.
Nuclear Regulatory Commission, "Assessment of the Nuclear Energy Institute 16-06,
'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis," May 30, 2017 (ML17031A269).
[15] J. Cushing, U.S. Nuclear Regulatory Commission, letter to J. V. Parrish, (Energy Northwest), "Review of Columbia Generating Station Individual Plant Examination of External Events Submittal (Tac No. M83695)," February 26, 2001, (ML010570035).
[16] NUREG-75/08, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants, LWR [Light Water Reactor] edition, December 1975.
[17]
[18]
NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991.
ASME Code Case, N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1," July 23, 2002.
Principal Contributor(s): Adrienne Brown John Hughey J. S. Hyslop Naeem Iqbal Alissa Neuhausen Keith Tetter Date: December 15, 2022
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