ML20218A660

Proposed Alternative I4R-01 to the Requirements of the ASME Code

ML20218A660
Person / Time
Site:	Clinton
Issue date:	09/02/2020
From:	Nancy Salgado Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Wiebe J
References
EPID L-2019-LLR-0112
Download: ML20218A660 (12)

Text

September 2, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

CLINTON POWER STATION, UNIT 1 - PROPOSED ALTERNATIVE I4R-01 TO THE REQUIREMENTS OF THE ASME CODE (EPID L-2019-LLR-0112)

Dear Mr. Hanson:

By letter dated December 16, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19350C642), as supplemented by letter dated May 13, 2020 (ADAMS Accession No. ML20134H998), Exelon Generation Company, LLC (the licensee) submitted request I4R-01 1 to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, requirements at Clinton Power Station (CPS), Unit 1.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed an alternative inservice inspection (ISI) program for Class 1 and 2 piping welds on the basis that the alternative provides an acceptable level of quality and safety.

The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the five key principles of risk-informed decision making are ensured by the licensee's proposed fourth 10-year risk-informed (RI)-ISI program. Therefore, the licensee's proposed fourth 10-year RI-ISI program is acceptable. The NRC staff finds that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC authorizes the use of the alternative RI-ISI program at CPS, Unit 1, for the fourth 10-year ISI interval, which commenced on July 1, 2020, and is scheduled to end on June 30, 2030.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

1 Other requests contained within the licensees letter dated December 16, 2019, have been or will be addressed via separate correspondence.

B. Hanson If you have any questions, please contact the Senior Project Manager, Joel S. Wiebe, at (301) 415-6606 or Joel.Wiebe@nrc.gov.

Sincerely, Digitally signed by Nancy L. Nancy L. Salgado Date: 2020.09.02 Salgado 09:25:46 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-461

Enclosure:

Safety Evaluation cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE I4R-01 REGARDING RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR CLASS 1 AND 2 PIPING EXELON GENERATION COMPANY, LLC CLINTON POWER STATION, UNIT 1 DOCKET NO. 50-461

1.0 INTRODUCTION

By letter dated December 16, 2019 (Agencywide Document Management and Access System (ADAMS) Accession No. ML19350C642, as supplemented by letter dated May 13, 2020 (ADAMS Accession No. ML20134H998), Exelon Generation Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. Relief request I4R-01 pertains to use of a risk-informed inservice inspection (RI-ISI) program for Class 1 and 2 piping welds at the Clinton Power Station (CPS), Unit 1.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed an alternative ISI program for Class 1 and 2 piping welds on the basis that the alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), ISI Standards Requirement for Operating Plants, throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME Code that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of 50.55a and that are incorporated by reference in paragraph (a)(1)(ii) of 50.55a, to the extent practical within the limitations of design, geometry, and materials of construction of the components.

Pursuant to 10 CFR 50.55a(g)(4)(ii), Applicable ISI Code: Successive 120-Month Intervals, ISI examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (a) of this section 18 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in U.S.

Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.147 Inservice Inspection Enclosure

Code Case Acceptability, ASME Section XI, Division 1, when using ASME BPV Code,Section XI, or NRC RG 1.192, when using the ASME OM Code, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section. However, a licensee whose ISI interval commences during the 12 through 18-month period after June 3, 2020, may delay the update of their Appendix VIII program by up to 18 months after June 3, 2020. Alternatively, licensees may, at any time in their 120-month ISI interval, elect to use the Appendix VIII in the latest edition and addenda of the ASME Code incorporated by reference in paragraph (a) of this section, subject to any applicable conditions listed in paragraph (b) of this section. Licensees using this option must also use the same edition and addenda of Appendix I, Subarticle I-3200, as Appendix VIII, including any applicable conditions listed in paragraph (b) of 10 CFR 50.55a.

Pursuant to 10 CFR 50.55a(z), Alternatives to Codes and Standards Requirements, alternatives to the requirements of paragraphs (b) through (h) of 50.55a or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that: (1) Acceptable Level of Quality and Safety, the proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a Compensating Increase in Quality and Safety, compliance with the specified requirements of 50.55a would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Background

During the second and third 10-year ISI intervals, for CPS, Unit 1, the licensee implemented NRC-approved RI-ISI programs for Class 1 and 2 piping welds. The licensee developed the original RI-ISI program, which was approved by the NRC staff by letter dated April 8, 2002 (ADAMS Accession No. ML020800820), in accordance with the NRC-approved methodology of the Electric Power Research Institute (EPRI) Topical Report (TR)-112657, Revision B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure December 1999 (ADAMS Accession No. ML013470102). Subsequently, by letter dated December 22, 2010 (ADAMS Accession No. ML103360335), the NRC authorized the RI-ISI program for the third 10-year ISI interval for CPS, Unit 1.

In its letter dated December 16, 2019, the licensee submitted relief request I4R-01 proposing to implement the RI-ISI program in the fourth 10-year ISI interval for CPS, Unit 1. As required, the licensee periodically evaluated its RI-ISI program and updated it as necessary during previous 10-year ISI intervals (second and third). The outcome of these reviews and updates is the latest CPS, Unit 1 RI-ISI program that is the subject of the current relief request I4R-01.

3.2 Components Affected ASME Code Class 1 and 2 piping welds are affected. These are:

• Class 1 vessel nozzle-to-pipe dissimilar metal (DM) welds classified as Examination Category B-F, Item No. B5.10 and 5.20;

• Class 1 piping similar and DM welds classified as Examination Category B-J, Item No.

B9.11, B9.21, B9.31, B9.32, B9.40;

• Class 2 austenitic stainless steel or high alloy piping welds classified as Examination Category C-F-I, Item No. C5.11; and,

• Class 2 carbon or low alloy steel piping welds classified as Examination Category C-F-2, Item No. C5.51 and C5.81.

3.3 Applicable Code Edition and Addenda The code of record for the fourth 10-year ISI interval is the 2007 Edition through 2008 Addenda of the ASME Code.

3.4 Duration of Request The licensee submitted this request for the fourth 10-year ISI interval for CPS, Unit 1, which commenced on July 1, 2020, and is scheduled to end on June 30, 2030.

3.5 ASME Code Requirement The ASME Code requirements applicable to the ISI of Class 1 and 2 piping welds originate in Table IWB-2500-1 and Table IWC-2500-1 of Section XI, respectively:

Examination Category B-F and B-J in Table IWB-2500-1 require Class 1 piping welds be subjected to volumetric examination, surface examination, or both during successive 120-month (10-year) intervals. According to the above requirements, 100 percent of all nozzle-to-pipe DM welds in Examination Category B-F, and 25 percent of all piping welds with more than one-inch nominal diameter in Examination Category B-J, shall be inspected.

Examination Category C-F-I and C-F-2 in Table IWC-2500-1 require Class 2 piping welds be subjected to the volumetric examination, surface examination, or both during successive 120-month (10-year) intervals. According to above requirements, 7.5 percent of non-exempt piping welds in Examination Category C-F-I and C-F-2 shall be inspected.

3.6 Proposed Alternative The licensee proposed an alternative to the ASME Code requirements. The proposed alternative is to use a plant-specific RI-ISI program described in the letters dated December 16, 2019, and May 13, 2020 (i.e., to continue implementing an RI-ISI program for Class 1 and 2 piping welds) in the fourth 10-year ISI interval for CPS, Unit 1.

3.7 Basis for Use of Alternative The licensee stated that the fourth 10-year ISI interval RI-ISI program will be a continuation of the current RI-ISI program as a living program. No changes to the evaluation methodology as currently implemented under EPRI TR-112657, Revision B-A, are required as part of this

interval update. Additionally, two enhancements to the RI-ISI program from ASME Code Case N-578-1, Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1, will continue to be implemented.

In tables of the Attachment to relief request I4R-01, the licensee summarized the results of its risk impact assessment for the fourth 10-year ISI interval. Based on its assessment, the licensee identifies the change in risk from the ASME Code ISI program to the RI-ISI program is shown to be less than 1.00E-06 per year for delta core damage frequency (CDF) and less than 1.00E-07 per year for delta large early release frequency (LERF).

3.8 NRC Staff Evaluation The NRC staff has evaluated this request for alternative pursuant to 10 CFR 50.55a(z)(1). The NRC staff focuses on whether the proposed alternative provides an acceptable level of quality and safety.

NRC Guidelines Utilized for this Evaluation NRC RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256) provides guidance on the use of probabilistic risk assessment (PRA) findings and risk insights to support licensee requests for changes to a plants licensing basis. NRC RG 1.174, Revision 3, also defines an acceptable approach to analyzing and evaluating proposed licensing basis changes. The approach includes traditional engineering evaluations supported by insights derived from the use of PRA methods about the risk significance of the proposed changes. In implementing risk-informed decision making, the NRC expects licensing basis changes meet the acceptance guidelines and key principles of risk-informed regulation specified in NRC RG 1.174, Rev. 3. Directly relevant to NRC RG 1.174, Revision 3, are:

• NRC RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014);

• NRC RG 1.178, Revision 1, An approach for Plant-Specific Risk-Informed Decision Making for Inservice Inspection of Piping (ADAMS Accession No. ML032510128); and,

• NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 3.9.8, Revision 0, Standard Review Plan for the Review of Risk-Informed Inservice Inspection of Piping (ADAMS Accession No. ML032510135).

NRC RG 1.200, Revision 2, describes an approach to determine whether the technical adequacy of the PRA used to support a submittal is consistent with accepted practices.

NRC RG 1.178, Rev. 1, provides guidance for development, review, approval, and implementation of a plant-specific RI-ISI program. NRC RG 1.178, Revision 1, describes methods acceptable to the NRC for integrating insights from PRA techniques with traditional engineering analyses into ISI programs for piping. Incorporating risk insights into the programs can focus inspections on the more important locations and reduce personnel exposure, while at the same time maintaining or improving public health and safety. The NRC SRP provides

guidance for evaluating the licensees requests for changes to the licensing basis using risk insights.

As a basis for its proposed alternative, the licensee used the methodology of the NRC-approved EPRI TR-112657, Revision B-A, along with enhancements specified in ASME Code Case N-578-1 to develop the initial RI-ISI program (approved by the NRC on April 8, 2002, ADAMS Accession No. ML020800820). The EPRI TR provides technical guidance on an alternative for selecting and categorizing the risk significance of piping components for the purpose of developing an RI-ISI program. The guidance in NRC RG 1.174, Revision 3, and RG 1.178, Revision 1, defines an acceptable approach to analyzing and evaluating the licensees proposed licensing basis changes that are supported with risk information. As part of evaluating the proposed change to the ISI program, the licensee performed an engineering analysis (i.e.,

traditional engineering evaluation methods supported by insights derived from the use of PRA methods about the risk significance of the proposed changes) to demonstrate that the proposed changes are in conformance with the key principles of risk-informed regulation in NRC RG 1.174, Revision 3, and will not compromise defense in depth(DID) and safety margins. As part of the RI-ISI process, the licensee performed periodic performance evaluations of the RI-ISI program and updated it in accordance with NRC RG 1.174, Revision 3, and NRC RG 1.178, Revision 1.

The key principles of risk-informed regulation in NRC RG 1.174, Rev. 3, are as follows:

Principle 1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12, Specific Exemptions.)

Principle 2. The proposed change is consistent with a DID philosophy.

Principle 3. The proposed change maintains sufficient safety margins.

Principle 4. When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

Principle 5. The impact of the proposed change should be monitored using performance measurement strategies.

Review of Adequacy of Fulfillment of Key Principles of Risk-informed Regulation In evaluating this relief request, the NRC staff focuses on whether the licensees proposed RI-ISI program for the fourth 10-year ISI interval conforms to these five key principles of risk-informed regulation. The NRC staff findings are as follows.

3.8.1 Principle 1 - Evaluation Principle 1 states that the proposed change must meet current regulations unless it is explicitly related to a requested exemption or rule change. The regulation in 10 CFR 50.55a(z) states, in part, that the Director of the Office of Nuclear Reactor Regulation may authorize an alternative to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a. For an alternative to be authorized per 10 CFR 50.55a(z)(1), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. The proposed RI-ISI change is an alternative to the ASME Code, Section Xl, as may be requested under 10 CFR 50.55a(z).

The proposed change is an alternative to piping ISI requirements with regard to the number of inspections, locations of inspections, and methods of inspections.

The NRC staff determined that the licensee met Principle 1 of RG 1.174, Revision 3, because the proposed RI-ISI program is an alternative to the ASME Code ISI program, as may be requested for NRC authorization pursuant to 10 CFR 50.55a(z)(1).

3.8.2 Principle 2 - Evaluation Principle 2 states that the proposed change must be consistent with the DID philosophy. ISI is an integral part of DID. As part of the RI-ISI process, the risk significance categorization and the specification of the subsequent number and location of elements to inspect will maintain the basic intent of ISI to identify and repair flaws before pipe integrity is challenged. Therefore, although a reduction in the number of welds inspected is anticipated, if a licensee implements a RI-ISI program as described in EPRI TR-112657 and subject to the conditions specified in the NRC staff safety evaluation report (SER) (ADAMS Accession No. ML993190474), there will be reasonable assurance that the program will provide a substantive ongoing assessment of piping condition.

In accordance with RG 1.174, Revision 3, engineering analysis should evaluate whether the impact of the proposed RI-ISI program is consistent with the DID philosophy. One aspect of the engineering evaluations is to show that the fundamental safety principles on which the plant design was based are not compromised by the proposed change. The EPRI approach to maintain DID is to characterize the role a piping system plays in the DID design principle and to review the potential changes in piping system performance that could be conceivably brought about.

The NRC staff confirmed that as part of the RI-ISI process, the licensee performed a plant-specific engineering analysis according to the guidance in the NRC-approved EPRI TR-112657, Revision B-A. The licensee performed this by assessing susceptibility of each piping segment to a particular degradation mechanism that may be a precursor to leak or rupture; assessing consequence of failure of the segment independent of failure potential; and, determining the risk significant locations and the number of locations to inspect. The NRC staff notes that the intent of the ISI is maintained by safety-significance categorization and the specification of the subsequent number and location of elements to inspect. This is used to ensure that DID is maintained, as discussed in NRC-approved EPRI TR-112657, Revision B-A. Therefore, the NRC staff determined that the licensee met Principle 2 of RG 1.174, Revision 3, and the proposed change is consistent with DID philosophy.

3.8.3 Principle 3 - Evaluation Principle 3 states that the proposed change shall maintain sufficient safety margins. No changes to the evaluation of design basis accidents in the final safety analysis report are being made by the RI-ISI process. Therefore, the NRC staff determined that the licensee met Principle 3 of RG 1.174, Revision 3, and the proposed changes are consistent with maintaining sufficient safety margin.

3.8.4 Principle 4 - Evaluation Principle 4 states that when proposed licensing basis changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement

on safety goals for the operations of nuclear power plants. Redirecting inspections to more risk significant locations and adaption of inspection procedures to the most likely degradation mechanisms at the specified locations is expected to contribute to a reduction of risk. This reduction in risk will partially or fully offset any risk increase from discontinuing inspection at low risk significant locations. This determination requires an estimate of the change in risk due to the alternative method. The change in the risk estimate is dependent on the location of inspections in the proposed RI-ISI program compared to the location of inspections that would be performed using the requirements of the ASME Code,Section XI.

Risk Metrics RG 1.178, Revision 1, provides that any risk increases that might result from a proposed RI-ISI program and their cumulative effects be small and not exceed NRC safety goals. Risk metric limits imposed by the EPRI TR-112657 methodology ensure that the change in risk of implementing the RI-ISI program meets the recommendations of RG 1.174, Revision 3, and RG 1.178, Revision 1. The EPRI criterion and RG 1.174, Revision 3, specify that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively.

The EPRI TR-112657 methodology discusses four screening evaluations in order of increasing resource requirements, as follows: qualitative, bounding without credit for any increase in probability of detection, bounding with credit for increase in probability of detection, and a Markov model-based calculation. Each licensee may select any of the screening evaluations.

The screening evaluations investigate the change in risk due to the change in the number and location of ISI inspections. All four screening evaluations include the assumption that there is a negligible risk increase because of the discontinuation of inspections of piping segments in the low-risk categories.

The licensees fourth ISI interval methodology is the same program methodology as approved in the third ISI interval. The risk impact assessment completed as part of the initial baseline RI-ISI program was an implementation/transition check on the initial impact of converting from a traditional ASME Code,Section XI, program to the new RI-ISI methodology. For the fourth interval ISI update, there is no transition occurring between two different methodologies, but rather, the previously-approved RI-ISI methodology and evaluation will be maintained for the new interval. The initial methodology of the evaluation has not changed, and the change in risk was re-assessed by comparing the initial ASME Code,Section XI, 1989 Edition program prior to the licensees implementation of a RI-ISI program and the updated fourth ISI interval RI-ISI program. Based on the licensees risk impact assessment of the fourth ISI interval update, the change in risk from the pre-RI-ISI ASME Code,Section XI program to the fourth interval RI-ISI program was below the 1E-06/year and 1E-07/year acceptance criteria for CDF and LERF, respectively, in RG 1.174, Revision 3. The CDF and LERF values for CPS, Unit 1, are 7.95E-09/year and 1.36E-09/year, respectively. These values are well below the "very small" change in core damage frequency guideline of 1E-6/year and change in LERF of 1E-07/year specified in RG 1.174, Revision 3. These values also meet the similar acceptance guidelines in EPRI TR-112657, Revision B-A.

Technical Adequacy of PRA Principle 4 also requires demonstration of the technical adequacy of the PRA. As discussed in RG 1.178, Revision 1 and RG 1.200, Revision 2, an acceptable change in risk evaluation (and risk-ranking evaluation used to identify the most risk significant locations) requires the use of a PRA of appropriate technical quality that models the as-built and as-operated plant. RG 1.200,

Revision 2, endorses the ASME/American Nuclear Society (ANS) PRA standard. The NRC approved EPRI TR, EPRI-112657, Revision B-A, provides guidance on the minimum acceptable quality requirement for a PRA used to support a RI-ISI program.

The enclosure to the licensees alternative request contains the licensees summary of the technical adequacy assessment of CPS, Unit 1, full power internal events (FPIE) PRA model (CL117A) to support this alternative request for continuation of the CPS, Unit 1, RI-ISI program for another 10-year interval. The most recent update of the CPS, Unit 1, FPIE PRA model (CL117A) was completed in August 2018, as a result of a regularly schedule update to the previous PRA model (CL114A). Section 3.1 of the Enclosure provides the description of the licensees PRA maintenance and update process to ensure that the applicable PRA model remains an accurate reflection of the "as-built, as operated" status of the plants. The licensee stated that the PRA technical adequacy assessment is in accordance with the requirements of RG 1.200, Revision 2, and NRC approved EPRI guidance. Information and impacts regarding PRA technical gaps relative to the ASME/ANS PRA Standard, as documented in the CPS, Unit 1, Peer Reviews and associated open Fact and Observations (F&Os) to support this RI-ISI application, are provided in Enclosure Tables 3-1 and 3-2. The Gap Assessment performed by the licensee demonstrates that the CPS, Unit 1, PRA model is technically adequate. The sensitivity analysis incorporating model changes addressing the open F&Os show that the base model remains adequate for this application. Thus, the NRC staff finds the licensee has assessed the technical adequacy of its PRA in accordance with the guidance of RG 1.200, Revision 2, and the quality of the PRA is sufficient to support the proposed RI-ISI program.

The NRC staff determined that the impact on CDF and LERF due to the implementation of the RI-ISI program satisfies the acceptance guidelines specified in RG 1.174, Revision 3, and EPRI TR, EPRI-112657, Revision B-A. The NRC staff also determined that the licensee has assessed the technical adequacy of its PRA in accordance with RG 1.200, Revision 2, and the PRA is consistent with the quality guidelines in EPRI TR-1021467-A. Therefore, the NRC staff determined that the licensee met Principle 4 of RG 1.174, Revision 3.

3.8.5 Principle 5 - Evaluation Principle 5 of risk-informed decision making requires that the impact of the proposed change be monitored by using performance measurement strategies. The RI-ISI program is a living program and, as such, is subject to periodic reviews. The licensee indicates that the Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, Element Selection, and Risk Impact Assessment steps encompass the living program process applied to the RI-ISI program. The letter dated December 16, 2019, stated that the evaluation and ranking procedure for the fourth 10-year ISI interval remain unchanged, and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-112657, Revision B-A. These portions of the RI-ISI program have been and will continue to be re-evaluated as major revisions of the site PRA occur and modifications to plant configuration are made.

Therefore, the NRC staff determined that the licensees proposed alternative provides assurance the fifth principle is met.

Based on the above, the NRC staff determined that the proposed RI-ISI program for the fourth 10-year ISI interval meets the five key principles of risk-informed regulation and, therefore, provides an acceptable level of quality and safety.

4.0 CONCLUSION

Based on the discussion in this safety evaluation, the NRC staff determined that the five key principles of risk-informed decision making are ensured by the licensee's proposed fourth 10-year RI-ISI program. Therefore, the licensee's proposed fourth 10-year RI-ISI program is acceptable. The NRC staff finds that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC authorizes the use of the alternative RI-ISI program at CPS, Unit 1, for the fourth 10-year ISI interval, which commenced on July 1, 2020 and is scheduled to end on June 30, 2030.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable.

Principal contributors: I. Dozier S. Cumblidge Date: September 2, 2020

ML20218A660 *via e-mail OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DNRL/NPHP/BC*

NAME JWiebe SRohrer MMitchell*

DATE 08/20/2020 08/06/2020 07/19/2020 OFFICE NRR/DRA/ARCB/BC* NRR/DORL/LPL3/BC*

NAME JDozier (A)* NSalgado DATE 07/17/2020 09/02/2020