ML23143A120
| ML23143A120 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/27/2023 |
| From: | Jennifer Dixon-Herrity Plant Licensing Branch IV |
| To: | Schuetz R Energy Northwest |
| Chawla M | |
| References | |
| EPID L-2022-LLR-0071 | |
| Download: ML23143A120 (1) | |
Text
June 27, 2023 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 - REQUEST NUMBER 4ISI-11 TO THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE REGARDING ALTERNATE EXAMINATION OF REACTOR VESSEL WELDS (EPID L-2022-LLR-0071)
Dear Mr. Schuetz:
By letter dated October 5, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22278B097), Energy Northwest (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, for the reactor pressure vessel (RPV) circumferential shell weld examinations at Columbia Generating Station (Columbia). The licensees proposed alternative is identified as Request Number 4ISI-11. The licensees request for the use of this alternative was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), on the basis that the alternative would provide an acceptable level of quality and safety.
The proposed alternative would eliminate the requirement to inspect the RPV circumferential shell welds, except for the areas of intersection with the axial welds, for the units 20-year extended license term. The licensees proposed alternative addressed the specific guidance referenced in NRCs Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds (ML031110082).
The NRC staff concludes that implementation of the BWRVIP-05 methods, in lieu of the specific ASME Code,Section XI, Category B-A, Item No. B1.11 requirements for volumetric examination of the subject RPV circumferential shell welds, will provide an acceptable level of quality and safety at Columbia for the units 20-year extended license term. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1).
Therefore, pursuant to 10 CFR 50.55a(z)(1), Request Number 4ISI-11 is authorized for the term of the Columbia renewed operating license, which ends on December 20, 2043.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.
R. Schuetz If you have any questions regarding this matter, please contact the Columbia Project Manager, Mr. Mahesh Chawla at (301) 415-8371 or via email to Mahesh.Chawla@nrc.gov.
Sincerely, Digitally signed by Samson S. Samson S. Lee Date: 2023.06.27 Lee 14:19:08 -04'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
Safety Evaluation cc: Listserv
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST NUMBER 4ISI-11 ALTERNATIVE REQUIREMENTS FOR REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELD EXAMINATIONS ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By letter dated October 5, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22278B097), Energy Northwest (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, for the reactor pressure vessel (RPV) circumferential shell weld examinations at Columbia Generating Station (Columbia). The licensees proposed alternative is identified as Request Number 4ISI-11. The licensees request for the use of this alternative was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), on the basis that the alternative would provide an acceptable level of quality and safety.
The proposed alternative would eliminate the requirement to inspect the RPV circumferential shell welds, except for the areas of intersection with the axial welds, for the units 20-year extended license term, also referred to as the period of extended operation (PEO). The licensees proposed alternative addressed the specific guidance referenced in NRCs Generic Letter (GL) 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds (ML031110082). This specific guidance provides staff expectations and acceptance criteria for plant-specific applications for ASME Code alternatives to implement the Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) topical report BWRVIP-05, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (proprietary) probabilistic fracture mechanics (PFM) methodology in lieu of the subject RPV circumferential shell weld examinations. An alternative in accordance with BWRVIP-05 to the RPV circumferential shell welds examination requirements of ASME Code,Section XI, for the remaining period of the current operating license was approved for Columbia on June 1, 2005 (ML051310091).
Enclosure
2.0 REGULATORY EVALUATION
The inservice inspection (ISI) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements, except where specific relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i),
Impractical ISI requirements: Granting of relief.
Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), lnservice inspection standards requirement for operating plants, ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year ISI interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii), 18 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2).
GL 98-05 provided recommendations for licensees planning to request permanent relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential welds. The recommendations included the need for licensees to perform their required inspections of essentially 100 percent of all axial welds. These recommendations were only applicable to the remaining term of operation under the initial existing license. Topical report BWRVIP-74-A, BWR Vessel and Internals Project: BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal (package (ML031710354); Non-proprietary (ML031710343); and proprietary (ML031710349)), demonstrates that NRC-approved inspection, flaw evaluation and aging management programs of the BWR vessel for the current operating license are also adequate for a 20-year license renewal term. The NRCs final safety evaluation (SE) (ML012920549) for approved topical report BWRVIP-74-A specifies action items for license renewal applicants for the PEO .
2.1 Requirements Related to Neutron Fluence The NRC has established requirements in Appendix G, Fracture Toughness Requirements, to 10 CFR Part 50, in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The regulations in 10 CFR Part 50, Appendix G, require that the pressure-temperature (P-T) limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G, Fracture Toughness Criteria for Protection Against Failure, to Section XI of the ASME Code were used to generate the P-T limits. The regulations in 10 CFR Part 50, Appendix G, also require that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.
Table 1, Pressure and Temperature Requirements for the Reactor Pressure Vessel, of 10 CFR Part 50, Appendix G, provides the NRC staffs criteria for meeting the P-T limit requirements of the ASME Code,Section XI, appendix G, as well as the minimum temperature requirements of the rule during normal and pressure testing operations. In addition, the NRC staffs regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ML003740284), and NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, section 5.3.2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, dated March 2007 (ML070380185).
The P-T limit curve calculations are based, in part, on the reference nil-ductility temperature (RTNDT) for the material, as specified in ASME Code,Section XI, appendix G. The regulations in 10 CFR Part 50, Appendix G, require that RTNDT values for materials in the RPV beltline region be adjusted to account for the effects of neutron radiation. The guidance in RG 1.99 contains methodologies for calculating the adjusted RTNDT (ART) due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the adjustment in reference temperature caused by irradiation (RTNDT), and a margin term.
The RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99 or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99 or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the neutron fluence, and the calculational procedures. The guidance in RG 1.99 describes the methodology to be used in calculating the margin term.
Appendix H, Reactor Vessel Material Surveillance Program Requirements, to 10 CFR Part 50, provides the NRC staffs criteria for the design and implementation of RPV material surveillance programs for operating LWRs.
The guidance in RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ML010890301), describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence with respect to the General Design Criteria (GDC) contained in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. In consideration of the guidance set forth in RG 1.190, GDC 14, Reactor coolant pressure boundary; GDC 30, Quality of reactor coolant pressure boundary; and GDC 31, Fracture prevention of reactor coolant pressure boundary, are applicable.
GDC 14 requires the design, fabrication, erection, and testing of the RCPB to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30 requires, among other things, that components comprising the RCPB be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31 requires the design of the RCPB have sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions that the boundary behaves in a nonbrittle manner and that the probability of rapidly propagating fracture is minimized. The design is to consider service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining
the material properties; the effects of irradiation on material properties; residual, steady-state and transient stresses; and the size of flaws.
Fluence calculations for use in ART and P-T limit curve analyses are acceptable if they are performed with approved methodologies or with methods that are shown to conform to the guidance in RG 1.190.
3.0 TECHNICAL EVALUATION
3.1 Licensees Request Number 4ISI-11 3.1.1 ASME Code Requirements to Which the Alternative are Requested The ASME Code,Section XI, 2007 Edition through 2008 Addenda, table IWB-2500-1, Examination Category B-A, Item No. B1.11 requires a volumetric examination of all the RPV circumferential shell welds each ISI interval, to include volumetric examination of essentially 100 percent (i.e., greater than 90 percent) of the length of the welds.
3.1.2 Components for Which the Alternative is Requested Code Class: 1 Examination Category: B-A Item No: B1.11, RPV Circumferential Welds Weld Nos.: AA, AB, AC, AD Examination Method: Volumetric 3.1.3 Licensees Proposed Alternative to the ASME Code Section XI Pursuant to 10 CFR 50.55a(z)(1), in lieu of the ASME Code requirements stated in section 2.0 of this SE, the licensee proposes to eliminate the RPV circumferential weld examinations required by the Code through the PEO. The licensee proposes to continue performing volumetric examinations of essentially 100 percent of the RPV axial shell welds and approximately 2 to 3 percent of circumferential welds at their points of intersection with the RPV axial welds.
The 40-year license term will end on December 20, 2023, for Columbia. The Request Number 4ISI-11 was submitted to implement the BWRVIP-05 PFM methods in lieu of the subject RPV circumferential shell weld examination requirements for the PEO.
3.1.4 Licensees Basis for the Proposed Alternatives The licensee submitted Request Number 4ISI-11 in accordance with 10 CFR 50.55a(z)(1), on the basis that the proposed alternatives would provide an acceptable level of quality and safety.
The licensees technical basis for determining an acceptable level of quality and safety included plant-specific evaluations for demonstrating that the limiting RPV circumferential shell weld at Columbia has conditional failure probabilities that are bounded by (i.e., less than) the NRC staffs acceptance criteria for the weld failure probabilities, considering projected RPV weld neutron embrittlement through the end of the PEO. The NRC staffs specific acceptance criteria for these circumferential shell weld failure probabilities were established in its final SE dated July 28, 1998, for the BWRVIP-05 report.
The licensee noted, as documented in the SE report (SER) for license renewal at Columbia (NUREG-2123, Volume 2 (ML12139A302)), section 4.2.5, Reactor Vessel Circumferential Weld Examination Relief, that the NRC staff previously found that Columbia submitted analyses for the PEO consistent with the evaluation criteria in the NRCs final SE for BWRVIP-05 but is still required to request relief in accordance with 10 CFR 50.55a for ISI intervals over the PEO.
The licensee determined that these RPV circumferential shell weld evaluations demonstrate that implementation of the proposed Code alternative for the 20-year extended license term would provide an acceptable level of quality and safety at Columbia.
3.2 NRC Staff Evaluation In September 1995, the BWR Owners Group submitted topical report BWRVIP-05, via Electric Power Research Institute (EPRI) (ML20098B288). The topical report provides the technical basis for eliminating inspections of RPV circumferential shell welds in BWR plants. In its review of BWRVIP-05, the NRC staff performed an independent PFM analysis to estimate RPV failure probabilities. By letter and associated SE dated July 28, 1998, the NRC staff approved the generic use of BWRVIP-05 and concluded that BWR licensees may request relief from examination of RPV circumferential shell welds by satisfying the following two criteria:
At the expiration of the operating license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998, safety evaluation.
Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staffs July 28, 1998, safety evaluation.
GL 98-05 permits BWR licensees to request permanent relief from the examination of RPV circumferential welds, provided that licensees comply with the above two criteria.
By letter dated June 1, 2005, the NRC staff authorized an alternative to the volumetric examination requirements of the ASME Code,Section XI, for the subject RPV circumferential shell welds at Columbia, pursuant to 10 CFR 50.55a(a)(3), which is now 10 CFR 50.55a(z)(1).
This NRC-authorized alternative allowed for plant-specific implementation of the BWRVIP-05 RPV PFM analyses, as approved by the NRC in its BWRVIP-05 final SE, in lieu of the subject ASME Code,Section XI examination requirements, for the duration of the units 40-year license term. The subject Code alternative will expire when Columbia enters the 20-year extended license term.
BWRVIP-74-A provides a technical basis for the elimination of RPV circumferential weld examination for the PEO. Item 11 in section 4.1, Renewal Applicant Action Items, of the NRC staff SE of BWRVIP-74-A states that a licensee may obtain relief from the ISI of the circumferential welds during the PEO by demonstrating that (1) at the end of the renewal period, the circumferential welds will satisfy the limiting conditional failure frequency for circumferential welds in the Appendix E of the staffs July 28, 1998 FSER [final SER] [on BWRVIP-05] and (2) that they have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staffs FSER.
By letter dated January 19, 2010 (ML100250656), the licensee submitted its license renewal application for the PEO (i.e., 54 effective full power years (EFPY)) at Columbia. Section 4.2.5,
Reactor Vessel Circumferential Weld Inspection Relief, of the Columbia license renewal application (ML100250654) discusses the time limiting aging analysis (TLAA) associated with the elimination of the RPV circumferential weld examination. In May 2012, the NRC published its SER approving the Columbias license renewal application in NUREG-2123. Section 4.2.5 of NUREG-2123, Volume 2, concludes that the licensees TLAA for Columbias RPV circumferential welds is acceptable for the PEO because (1) the conditional failure probability for 54 EFPY of operation will remain bounded by the NRCs analysis in the its final SE for BWRVIP-05, (2) the licensee will be using procedures and training to limit cold overpressure events during the PEO, and (3) the licensees TLAA is consistent with the evaluation criteria in the NRCs final SE for BWRVIP-05.
The licensees submittal dated October 5, 2022, requested alternatives to the subject circumferential weld examination requirements for the PEO at Columbia based on plant-specific implementation of the NRC-approved BWRVIP-05 methods for the limiting circumferential shell weld at Columbia, considering projected RPV weld neutron embrittlement through 60 years of facility operation. The proposed 60-year Code alternatives included plant-specific calculations demonstrating that projected neutron embrittlement for the Columbia limiting RPV circumferential shell weld is less than that used by the NRC staff for calculating an acceptable circumferential shell weld conditional failure probability,1 as documented in the NRCs final SE for BWRVIP-05.
The specific RPV weld neutron embrittlement parameter used for this evaluation is referred to as the mean RTNDT. The mean RTNDT value for demonstrating an acceptable RPV weld conditional failure probability is calculated based on three inputs:
(1) The Projected RPV Neutron Fluence: RPV neutron fluence, as determined based on staff-approved calculation methodologies, is the key time-dependent parameter for all RPV integrity analyses that consider neutron embrittlement of the RPV beltline materials.
The projected neutron fluence input to the mean RTNDT value, for demonstrating an acceptable RPV weld conditional failure probability at the end of the licensed operating term, shall include the effects of any power uprates that are implemented during the licensed operating term of the unit.
(2) The Weld CF: The CF is determined based on both copper and nickel content, or the application of credible RPV material surveillance data from a 10 CFR Part 50, Appendix H RPV material surveillance program. If the weld is represented in the plant-specific or industry integrated surveillance program, all credible RPV surveillance data shall be used for the CF calculation, per the requirements of 10 CFR Part 50, Appendix G. CF values shall be periodically recalculated based on new credible RPV surveillance data that becomes available when a surveillance capsule is withdrawn from the RPV and tested in accordance with 10 CFR Part 50, Appendix H surveillance program requirements.
(3) The Initial (Unirradiated) RTNDT: The initial RTNDT is determined in accordance with the requirements of 10 CFR Part 50, Appendix G, based on the procured RPV material impact test data or the use of NRC-approved methods in NUREG-0800, Branch Technical Position 5-3, Revision 2, Fracture Toughness Requirements, dated March 2007 (ML070850035), as applicable to the unit. This is expected to remain fixed 1
The weld conditional failure probability quantifies the probability of weld failure if the RPV was subjected to a cold overpressure event, as addressed in BWRVIP-05.
throughout the operating life of the plant.
The NRC staff evaluated Request Number 4ISI-11 with respect to the NRC staffs SE for BWRVIP-05 to determine whether the proposed request complies with the above two criteria.
Criterion 1 - Conditional Failure Probability of Columbia RPV Circumferential Welds The BWRVIP-05 report states that the ART value is the parameter that is used to assess the degree of embrittlement in RPV circumferential welds affected by neutron fluence irradiation.
According to the NRCs final SE for BWRVIP-05, the NRC staff independently calculated generic mean ART values of welds in the RPV fabricated by various vendors per RG 1.99. The NRC staff compared its calculated generic mean ART values with the ART values calculated by BWRVIP-05. The NRC staff noted that the ART values are used to calculate the conditional probability of failure of RPV circumferential welds. The NRC staff compared its calculated conditional probabilities of failure of the welds with that of BWRVIP-05. The NRCs final SE for BWRVIP-05 concludes that the conditional probability of failure for site-specific RPV circumferential welds is adequately bounded and acceptable if the limiting mean ART of the site-specific RPV circumferential welds is less than or equal to the generic mean ART calculated by BWRVIP-05.
In response to Criterion 1, the licensee discussed the key parameters that affect the mean ART of the subject RPV circumferential welds during plant operation and, in turn, affect the conditional failure probability of the circumferential welds. These key parameters are neutron fluence, weld chemistry, initial RTNDT and RTNDT. The licensee derived these key parameters for the limiting RPV circumferential weld at Columbia and compared them to the corresponding parameters used for the generic circumferential weld in the NRCs final SE for BWRVIP-05. The licensees comparison is shown in table 2 of request number 4ISI-11, which states that Columbias RPV is fabricated by the Chicago Bridge and Iron Company (CB&I). The licensee compared the Columbias key parameters to the key parameters for the generic circumferential weld in CB&I fabricated RPV in the NRCs final SE for BWRVIP-05.
In Request Number 4ISI-11, the licensee states that the circumferential weld AB with Heat No. 5P6756 located between shell course 1 and shell course 2 within the beltline region as the limiting circumferential weld among the four affected welds. As shown in table 2 of Request Number 4ISI-11, for the limiting mean ART of the Columbia RPV circumferential weld AB, the licensee used or calculated (1) an initial RTNDT of minus (-) 50 degrees Fahrenheit (°F),
(2) a limiting neutron fluence value of 4.20E17 neutron/cm2 (n/cm2) at 54 EFPY, (3) a CF of 108ºF, and (4) a RTNDT value of 33ºF. Based on these key parameters, the licensee calculated a limiting mean ART of -17ºF for the Columbia RPV circumferential weld AB.
In the NRCs final SE of BWRVIP-05 for the assessment of the generic, limiting CB&I fabricated RPV circumferential weld, the NRC staff applied a limiting end-of-license neutron fluence of 1.02E19 n/cm2 for the mean ART at 64 EFPY as shown in table 2 of Request Number 4ISI-11 and table 2.6-5 of the NRCs final SE for BWRVIP-05. This neutron fluence resulted in a maximum NRC staff-approved upper bound limiting mean ART value of 70.6ºF for the generic CB&I fabricated RPV circumferential weld as shown in table 2 of Request Number 4ISI-11.
In comparison, the NRC staff noted that the Columbia-calculated mean ART for its limiting circumferential weld AB of -17°F is bounded by the limiting mean ART of generic circumferential weld of 70.6°F. The NRC staff observes that the limiting mean ART of -17ºF reported for circumferential weld AB represents a slight change to the prior mean ART of -6ºF for this weld in
section 4.2.5 of NUREG-2132, Volume 2, where the value of -6ºF was calculated based on a limiting projected neutron fluence of 4.78E17 n/cm2. However, the NRC staff finds that the limiting mean ART value of -17ºF for circumferential weld AB to be acceptable based on the revised projected neutron fluence of 4.20E17 n/cm2 that was: (1) reported for this weld in a license amendment request dated October 13, 2021 (ML21299A182), and (2) approved in the NRC staffs SE for the associated license amendment request (License Amendment No. 268) dated November 23, 2022 (ML22263A445). In the NRC staff SE associated with Amendment No. 268, the NRC staff found that the proposed P-T curves are based on acceptable fluence inputs because the fluence calculational methodology has been approved for use by the NRC staff for facilities like Columbia and it adheres to the guidance in RG 1.190.
Based on the assessment above, the NRC staff determined that margin exists in the mean ART value for the Columbia RPV circumferential welds (represented by limiting circumferential weld AB) at 54 EFPY as compared to the corresponding mean ART value of the generic CB&I fabricated RPV circumferential weld in the NRCs final SE for BWRVIP-05. As such, the NRC staff finds that the licensee has demonstrated that the conditional failure probability of the Columbias circumferential welds is bounded by the NRC staffs PFM analysis results as shown in the NRCs final SE for BWRVIP-05. Therefore, the NRC staff concludes that the mean ART calculation for the subject Columbia circumferential welds satisfies Criterion 1 of the NRCs final SE for BWRVIP-05 and that the conditional probability of failure for the Columbia RPV circumferential welds is bounded by the corresponding conditional probability of failure for generic CB&I fabricated RPV circumferential welds in the NRCs final SE for BWRVIP-05.
Based on its review of the parameters relevant to the proposed request and comparison with respect to the 64 EFPY analyses in the NRCs final SE for BWRVIP-05, the NRC staff concludes that the Columbia circumferential welds continue to satisfy the limiting conditional failure probability for circumferential welds in Appendix E of the NRCs final SE for BWRVIP-5 and satisfied GL 98-05.
Criterion 2 - Operator Training and Established Procedures of Cold Over-Pressure Events The NRC staff confirmed that the proposed 60-year Code alternatives continued implementation of certain operator procedures and training needed to limit the frequency of cold overpressure events, per the criteria specified in the NRCs final SEs for BWRVIP-05 and BWRVIP-74-A. The NRC staff had previously endorsed these provisions in section 4.2.5 of its SER for the Columbia license renewal application in NUREG-2123, regarding the subject circumferential weld analyses. Additionally, the NRC staff finds that the description of operator procedures and training in the subject request is substantially unchanged from those described during the initial licensing term as discussed in the relief request dated July 15, 2004 (ML042150393).
Procedural controls will continue to include continuous monitoring of indications and alarms during cold shutdown and refueling operations, requirements to perform certain changes under the auspices of a senior reactor operator, and shift briefs to ensure situational awareness of on-coming operators. Operator training will continue to include elements such as brittle fracture and vessel thermal stress concepts; relevant operational limits; surveillance procedures to ensure limits are adhered to; operational transient procedures; simulator training; and expectations for strict procedural compliance. Additionally, the licensee will continue to perform work management activities (such as review of planned outage work activities against a shutdown safety plan, coordination, and oversight of outage activities through an outage control center, and mandatory pre-job briefings) and regularly review industry operating experience to ensure that Columbias procedures and training are updated in keeping with lessons learned at other plants. The NRC staff finds that these activities are sufficient to provide reasonable assurance
that the overall plant-specific RPV failure probability per reactor operating year (a product of the weld conditional failure probability and the cold overpressure event frequency) is less than the acceptance criterion specified in the NRCs final SE for BWRVIP-05, which satisfies Criterion 2.
4.0 CONCLUSION
The NRC staff finds that the information submitted by the licensee demonstrates that the conditional failure probabilities for the Columbia limiting RPV circumferential shell welds at the end of the PEO satisfies the NRC staffs acceptance criteria for these evaluations in its final SE for BWRVIP-05, GDC 14, GDC 30 and GDC 31. Additionally, the licensee will continue to implement operator training and procedures to limit the frequency of cold overpressure events in accordance with the NRCs final SE for the BWRVIP-05 report, consistent with the staffs previous approval of these methods for the PEO, as documented in section 4.2.5 of NUREG-2123 for the license renewal of Columbia. The licensee has therefore satisfied the plant-specific conditions required to obtain NRC authorization for this specific Code alternative.
On this basis, the NRC staff concludes that implementation of the BWRVIP-05 methods, in lieu of the specific ASME Code,Section XI, Category B-A, Item No. B1.11 requirements for volumetric examination of the subject RPV circumferential shell welds, will provide an acceptable level of quality and safety at Columbia for the units 20-year extended license term.
The NRC staff has reviewed the subject request and concludes as set forth above, that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, pursuant to 10 CFR 50.55a(z)(1), Request Number 4ISI-11 is authorized for the term of the Columbia renewed operating license, which ends on December 20, 2043.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributors: A. Rau, NRR J. Tsao, NRR J. Medoff, NRR Date: July 27, 2023
ML23143A120 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DNRL/NVIB/BC NAME MChawla PBlechman ABuford (OYee for)
DATE 5/22/2023 5/24/2023 6/2/2023 OFFICE NRR/DSS/SNSB/BC* NRR/DORL/LPL4/BC NAME PSahd (SBhatt for) JDixon-Herrity (SLee for)
DATE 6/7/2023 6/27/2023