ML14321A864

Relief Requests 13R-08, Reactor Pressure Vessel (RPV) Interior Attachments and 13R-09, RPV Pressure-Retaining Welds, Exam Interval Extensions, Third 10-year Inservice Inspection Interval (TAC Nos. MF3321 and MF3322)

ML14321A864
Person / Time
Site:	Wolf Creek
Issue date:	12/10/2014
From:	Oesterle E R Plant Licensing Branch IV
To:	Heflin A C Wolf Creek
Lyon C F
References
TAC MF3321, TAC MF3322
Download: ML14321A864 (13)

Text

Mr. Adam C. Heflin UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 10, 2014 President, Chief Executive Officer, and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF C,REEK GENERATING STATION-REQUEST FOR RELIEF NOS. 13R-08 AND 13R-09 FOR THE THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (TAC NOS. MF3321 AND MF3322)

Dear Mr. Heflin:

By letter dated January 8, 2014, as supplemented by letter dated August 28, 2014, Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) proposed alternatives to the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, lSI Program, for the Wolf Creek Generating Station (WCGS). Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(a)(3)(ii), relief request 13R-08 proposed an extension of the interval by approximately 3 months beyond the Code-allowed end-of-interval extension for the performance of ASME Code-required Category B-N-2 and B-N-3 examinations of the reactor pressure vessel (RPV) interior attachments and core support structure.

Pursuant to 10 CFR 50.55a(a)(3)(i), relief request 13R-09 proposed an extension of the interval from 10 to 20 years for the performance of ASME Code-required Category B-A (volumetric examinations of the RPV welds) and Category B-D (the nozzle inside radius and nozzle weld). The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject requests and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed

<:!II of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) for 13R-08 and 10 CFR 50.55a(a)(3)(i) for 13R-09. Therefore, the NRC authorizes the use of relief request 13R-08 at WCGS until 3 months beyond the end of the current third lSI interval, which wil! be the conclusion of refueling outage RF21, currently scheduled for the fall of 2016. The NRC also authorizes the use of relief request 13R-09 for Category B-A (volumetric examinations of the RPV welds) and Category B-D (the nozzle inside radius and nozzle weld) at WCGS until the end of the extended third lSI interval in 2025. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

A Heflin The detailed results of the NRC staff review are provided in the enclosed safety evaluation.

If you have any questions concerning this matter, please call Mr. F. Lyon of my staff at (301) 415-2296 or by electronic mail at fred.lyon@nrc.gov.

Docket No. 50-482 Enclosure Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL REQUEST FOR RELIEF NOS. 13R-08 AND 13R;09 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482

1.0 INTRODUCTION

By letter dated January 8, 2014, as supplemented by letter dated August 28, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 14014A099 and ML 14251A030, respectively), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) proposed alternatives to the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, lSI Program, for the Wolf Creek Generating Station (WCGS). Attachment 1 of the January 8, 2014, letter contains relief request (RR) 13R-08. Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(a)(3)(ii), this alternative requests an extension of the interval by approximately 3 months beyond the Code-allowed of-interval extension for the performance of ASME Code-required Category B-N-2 and 8-N-3 examinations of the reactor pressure vessel (RPV) interior attachments and core support structure.

Attachment 2 of the January 8, 2014, letter contains RR 13R-09. Pursuant to 10 CFR 50.55a(a)(3)(i), this alternative requests an extension of the interval from 10 to 20 years for the performance of ASME Code-required Category 8-A (volumetric examinations of the RPV welds) and Category 8-D (the nozzle inside radius and nozzle weld). This request is based on the methodology defined in Westinghouse topical report WCAP-16168-NP-A, Revision 3, and is consistent with the latest industry implementation plan, Pressurized Water Reactor Owners Group (PWROG) OG-1 0-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval perWCAP-16168-NP, Revision 1, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,"'

dated July 12, 2010 (ADAMS Accession No. ML 11153A033).

The current third 1 0-year lSI interval ends on September 2, 2015, for the WCGS RPV. Enclosure

2.0 REGULATORY EVALUATION

2.1 Regulations

and Guidance In accordance with 10 CFR 50.55a(g)(4), the licensee is required to perform lSI of ASME Code Class 1, 2, and 3 components and system pressure tests during the first 1 0-year interval and subsequent 1 0-year intervals that 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(b), subject to the limitations and modifications listed therein. For the third 1 0-year lSI intervals at WCGS, the Code of record for the inspection of ASME Code Class 1, 2, and 3 components is the 1998 Edition through the 2000 Addenda of the ASME Code,Section XI. The regulation in 10 CFR 50.55a(a)(3) states, in part, that the Director '--of the Office of Nuclear Reactor Regulation may authorize an alternative to the requirements of 10 CFR 50.55a(g).

There are two justifications for an alternative to be authorized.

First, per 10 CFR 50.55a(a)(3)(i), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. For the second possible justification-for an alternative to be authorized, described in 10 CFR 50.55a(a)(3)(ii), the licensee must show that following the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (ADAMS Accession No. ML003740284), describes general procedures acceptable to the staff for calculating the effects of neutron radiation embrittlement of the alloy steels currently used for light-water-cooled RPVs. RG 1.174, Rev. 1, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002 (ADAMS Accession No. ML023240437), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.

• RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron I Fluence," March 2001 (ADAMS Accession No. ML011500010), describes methods and assumptions acceptable to the staff for determining the RPV neutron.fluence.

2.2 Background

The lSI of Categories B-N-2, B-N-3, B-A, and B-D components consists of visual and/or ultrasonic examinations intended to discover whether flaws have initiated, whether pre-existing . flaws have extended, and whether pre-existing flaws may have been missed in prior examinations.

These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code. 2.3 Summary ofWCAP-16168-NP-A.

Rev. 2 In June 2008, the PWROG issued the NRC-approved topical report WCAP-16168-NP-A, Rev. 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (ADAMS Accession No. ML082820046), which is in support of a risk-informed assessment of extensions to the lSI intervals for Categories'B-A and B-D components.

Specifically, WCAP-16168-NP-A, Rev. 2 took data associated with three different pressurized-water reactor (PWR) plants (referred to as the pilot plants), one designed by each of the three main vendors (Westinghouse, Combustion Engineering (CE), and Babcock and Wilcox (B&W)) for PWR nuclear power plants in the United States, and performed studies on these pilot plants to justify the proposed extension of the lSI interval for Categories B-A and B-0 components from 10 to 20 years. The analyses in WCAP-16168-NP-A, Rev. 2 used probabilistic fracture mechanics (PFM) tools and inputs from the work described in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)," August 2007 (ADAMS Accession No. ML072830016), and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (publicly available at rm/doc-collections/nuregs/staff/sr1874/sr1874.pdf ADAMS Accession No. ML070860156).

The PWROG analyses incorporated the effects of fatigue crack growth and lSI. Design basis transient data was used as input to the fatigue crack growth evaluation.

The effects of lSI were modeled consistent with a previously-approved PFM Code in WCAP-14572-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection" (ADAMS Package Accession No. ML012630375).

These effects were considered in the PFM evaluations, using the Fracture Analysis of Vessels-Oak Ridge (FAVOR) computer code (ADAMS Accession No. ML042960391

). All other inputs were identical to those used in the*PTSrisk re-evaluation underlying 10 CFR 50.61a. From the results of the studies, the PWROG concluded that the ASME Code,Section XI 1 0-year inspection interval for Categories B-A and B-D components in PWR RPVs can be extended to 20 years. Their conclusion from the results for* the pilot plants was considered to apply to any plant designed by the three vendors, as long as the critical, plant-specific parameters (defined in Appendix A of WCAP-16168-NP-A, Rev. 2) are bounded by the pilot plants.

• 2.4 Summary of the July 26. 2011. NRC Safety Evaluation for WCAP-16168-NP-A, Rev. 2 The original NRC safety evaluation (SE) in WCAP-16168-NP-A, Rev. 2, published in 2008, was superseded by an SE dated July 26, 2011 (ADAMS Accession No. ML 111600303), to address . the PWROG's request for clarification of the information needed in applications utilizing WCAP-16168-NP-A,

2. The staff's conclusion in this latter SE indicates that the

• methodology presented in WCAP-16168-NP-A, Rev. 2 is consistent with RG 1.174, Rev. 1 and is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and conditions in the SE. In addition to showing that the subject plant parameters and inspection history are bounded by the critical parameters identified in Appendix A in WCAP-16168-NP-A, Rev. 2, the licensee's application must provide the following plant-specific information:

(1) Licensees must demonstrate that the embrittlement of their RPV is within the envelope used in the supporting analyses.

Licensees must provide the 95th percentile total through-wall cracking frequency (TWCFrorAd and it's supporting material properties at the end of the period in which the relief is requested to extend the lSI from 10 to 20 years. The 95 1 h percentile TWCFroTAL must be calculated using the methodology in NUREG-1874.

The RT MAx-x and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, LlTJo, must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRC-approved methodology.

(2) Licensees must report whether the frequency of the limiting design basis transients during prjor plant operation are less than the frequency of the design basis transients identified in the PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth. (3) Licensees must report the results of prior lSI of RPV welds and the proposed schedule for the next 20-year lSi interval.

The 20-year inspection interval is a maximum interval.

In its request for an alternative, each licensee shall identify the years in which future inspections will be performed.

The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation pla*n provided to the NRC in PWROG letter OG-1 0-238 (ADAMS Accession No. ML 11153A033).

(4) Licensees with 8&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all of its design basis transients and (b) identify the design bases transients that contribute to significant fatigue crack growth. (5) Licensees with RPVs having forgings that are susceptible to underclad cracking and with RT MAX-Fa values 240 degrees Fahrenheit CF) must submit a plant-specific evaluation to extend the inspection interval for ASME Code,Section XI, Category 8-A and 8-D RPV welds from 1 0 to a maximum of 20 years because the analyses performed in the WCAP-A are not applicable.

(6) Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61 a. WCAP-16168-NP-A, Rev. 3, which contains this latter SE for WCAP-16168-NP-A, Rev. 2, was issued in October 2011 (ADAMS Accession No. ML 11306A084, referred to as the WCAP-A in the rest of this SE).' 3.0 TECHNICAL EVALUATION 3.1 RR 13R-08 3.1.1 Description of Proposed Alternative In RR 13R-08, the licensee proposed extending the current inspection interval for the Category 8-N-2 and 8-N-3, visual testing (VT-3) of core support structure from the end of RF 20 to RF 21, the next scheduled outage where the core barrel and fuel will be removed. 3.1.2 Components for Which Relief is Requested

\ The affected components are the subject plant RPVs and their interior attachments and core support structures.

The following examination categories and item numbers from IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI, are addressed in this request: For relief request 13R-08: Exam Category

• B-N-2 Item Number B13.60 B13.70 3.1.3 Reason for Proposed Alternative Description Interior Attachments Beyond Beltline Region Removable Core Support Structures

_ These inspections are typically done at the end of an interval in conjunction with the ultrasonic testing (UT) of the RPV welds, which is normally the only time when there is a full offload of the core and the core barrel, allowing access to the core support structure.

For the third interval at WCGS, this corresponds to Refueling Outage 20 (RF20). In RR 13R-09, the licensee has requested to extend the third interval from 1 0 to 20 years for the UT of the RPV welds so that the UT will not be done again until 2025, plus or minus one refueling outage. Given the extension of the third interval for the UT of the RPV welds, the core support structures will not be accessible for the required VT-3 examinations.

3.1.4 Proposed

Alternative and Basis for Use Performing the Code-required B-N-2 and B-N-3 inspections requires removal of the core barrel, but to remove the core barrel for the sole purpose of performing these inspections represents an unnecessary risk and would increase the radiation exposure to inspectors if the examinations are not combined with other required maintenance.

Therefore, an extension is requested from the Code requirements so that the VT-3 of the core support structures can be performed at the opportunity, which will be the next outage (RF21) when a full core offload and core barrel removal is scheduled to allow for examination and mitigation activities on the RPV hot and cold leg nozzle welds at WCGS: This proposed alternative does not change the ASME Category B-N-1 inspections.

These visual examinations of the space above and below the core made accessible during normal refueling outages will be performed as required by the Code. In accordance with 10 CFR 50.55a{a){3)(ii), this 3 month extension of the B-N-2 and B-N-3 inspections is requested on the basis that compliance with the Code required inspections of the core support structures would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. 3.1.5 Duration of Proposed Alternative ASME Code,Section XI, IWA-2430(d)(1) allows inspection intervals to be extended by up to 1 year (12 months). The proposed alternative would perform the required B-N-2 and B-N-3 inspections during RF21, which will require a 15-month extension of the third interval, 3 months more than allowed by regulation.

3.1.6 Precedents

The licensee noted that this same alternative has been granted by the NRC at two other PWRs in the United States. 3.2 RR 13R-09 3.2.1 Description of Proposed Alternative In RR 13R-09, the licensee proposed extending the current inspection interval from 10 to 20 years, deferring the ASME Code required Categories 8-A and 8-D weld lSI for WCGS until 2025. This schedule is consistent with the schedule proposed in the revision to PWROG Letter OG-10-238.

3.2.2 Components

for Which Relief is Requested The affected components are the subject plant's RPVs and their interior attachments and core support structures.

The following examination categories and item numbers from IW8-2500 and Table IW8-2500-1 of the ASME Code,Section XI, are addressed in this request: For RR 13R-09: Exam Category 8-A 8-A 8-A '8-A 8-A 8-A 8-D 8-D Item Number 81.11 81.12 81.21 81.22 81.30 81.40 83.90 83 .. 100 3.2.3 Reason for Proposed Alternative Description Circumferential Shell Welds Longitudinal Shell Welds Circumferential Head Welds Meridional Shell Welds Shell-to-Flange Weld Head-to-Flange Weld Nozzle-to-Vessel Welds Nozzle Inner Radius Section The licensee stated that the proposed alternative will reduce radiation exposure to inspectors and the cost ofCode-required examinations. 3.2.4 Basis for Proposed Alternative The licensee stated that the methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Category 8-A and B-D components is contained in the WCAP-A. This methodology used the estimated TWCF as a measure of the risk of RPV failure, and it was demonstrated that the inspection interval for the affected components can be extended from 10 to 20 years, meeting the change in risk guidelines in RG 1.174. The licensee addressed the plant-specific information discussed in Section 2.4 of this SE as follows: (1) A plant-specific analysis, with identified critical parameters and detailed TWCF calculations demonstrated that the WCGS RPVs' parameters are bounded by corresponding pilot plant parameters.

The total TWCFs were calculated as 7.01E-14 for WCGS, less than the value of 1.76E-08 for the Westinghouse pilot plant in the WCAP-A. (2) The frequencies of the WCGS RPVs' limiting design basis transients are bounded by the frequencies identified in the PWROG fatigue analysis.

(3) The results of the previous RPV inspections for the WCGS RPV are provided (zero reportable indications found with last inspections with methodology based on ASME Code,Section XI and Section V, 1989 Edition and 1995 Addenda), which confirm that satisfactory examinations have been performed on the WCGS RPVs. Therefore, the licensee stated that the lSI results are acceptable per the requirements of 10 CFR 50.61 a. The licensee has not addressed plant-specific information items (4), (5), and (6) because they do not apply to the WCGS RPV. Since the plant-specific information for the WCGS RPV are bounded by the pilot plant application, the licensee concluded that use of this proposed alternative will provide an acceptable level of quality and safety and, therefore, pursuant to 10 CFR 50.55a(a)(3)(i), requested that the NRC authorize the relief. 3.2.5 Duration of Proposed Alternative The third 1 0-years lSI interval is scheduled to end in 2015. Granting of this proposed alternative will allow the interval to be extended and the UT inspections will be scheduled no later than 2025. 3.2.6 Precedents The licensee noted that this same alternative has been granted by the NRC at eight other PWRs in the United States. 3.3 NRC Staff Evaluation 3.3.1 RR 13R-08 The NRC staff has reviewed the information in Attachment 1 to the January 8, 2014, submittal.

Historically, 'the RPV welds and the VT-3 inspections of the core support structures have always been performed at the same time. The licensee's request to extend the RPV weld inspection (RR 13R-09) was the second part of the January 8, 2014, submittal.

Given approval of RR 13R-09, there is no reason the core barrel and fuel would be removed other than the lSI VT-3 inspections associated with this relief request. The staff notes that every time the core barrel and fuel are removed from the unit, there is a risk associated with that activity, and there will be additional radiation exposure to workers in the area. Therefore, to minimize risk and reduce radiation exposure of the workers to as low as is reasonably achievable (ALARA), the licensee has proposed this alternative.

The NRC staff agrees with the licensee's determination that removal of the core barrel and fuel for the sole purpose of performing the lSI VT-3 examinations of the core support structures so that the timing of the examinations would meet the ASME Code,Section XI lSI interval requirements would create a risk and represents a hardship that comes without a compensating increase in the level of quality and safety. The staff also agrees that the proposed alternative for WCGS is the same as has already been granted to other PWRs by-the NRC. Therefore, based on the above considerations, the staff concludes that the licensee's request to defer the lSI VT -3 examinations for the WCGS RPV from RF20 to RF21 (a total of 15 months, 3 additional months beyond that allowed in Section XI) is acceptable.

3.3.2 RR 13R-09 The NRC staff has reviewed the information in Attachment 2 to the January 8, 2014, submittal.

Tables 1, 2, and 3 of Attachment 2 include all of the plant-specific information required for application of the topical report to the WCGS RPV. The staff noted one issue related to the number of past inspections in Table 2 of Attachment 2; it appears that more than two inspections were done on some of the welds. To clarify the inspection history, by letter dated July 31, 2014, the NRC staff issued a request for additional information (RAI) asking for a more detailed description of what inspections were performed during the first two 1 0-year lSI intervals.

By letter dated August 28, 2014, the licensee responded to the RAI saying that only two inspections were done on each weld, one for each interval.

The appearance that more inspections were done is a result of partial examinations of the welds spread over the interval, which combined to provide one complete inspection of the Category 8-A flange welds. . Also, the WCGS RPV has a single-layer cladding on the inside like the assumption used in the WCAP-A analysis.

Therefore, the NRC staff determined that the licensee has addressed specific information item 2 (see SE Section 2.4) satisfactorily and confirmed that the issue raised in the RAI, regarding past inspection history, is resolved.

The TWCF calculations used inputs from Table 3 of Attachment 2 to the January 8, 2014, submittal.

The relief request used RG 1. 99, Rev. 2 to calculate ll T 30. The calculations used

• Position 1.1 (Table 2 of the regulatory guide) for five of the six plates and Position 2.1 (credible surveillance data available), for the one plate, which was the limiting material for each of the three inputs to the total TWCF, and all RPV beltline welds. The NRC staff independently calculated the TWCF and the results were in good agreement with that reported by the licensee.

The peak fluence values used in Table 3 of Attachment 2 to the January 8, 2014, submittal are about 5 percent less than those used in license renewal. This difference in peak fluence values is acceptable for this application because the licensee's calculated total TWCF of 7.01 E-14 is several orders of magnitude lower than the value of 1. 76E-08 for the Westinghouse pilot plant in the WCAP-A; the 5 percent higher peak fluence would have a negligible effect on the calculated total TWCF. The NRC staff confirmed that the chemistry values were adequately generated using surveillance data in accordance with RG 1.99, Rev. 2. Hence, the NRC staff determined thatthe licensee has addressed plant-specific information item 1 (see SE Section 2.4) satisfactorily and confirmed that the embrittlement of the WCGS RPV is bounded by that determined for the Westinghouse pilot plant in the topical report. ' Regarding the 10 CFR 50.61 a requirements on allowable flaws, the licensee noted that no reportable indications were found during past inspections and that therefore, the results are acceptable.

The NRC staff notes that no additional evaluation is required to use the WCAP-A. The next inspection for WCGS would be conducted in 2025. The NRC staff has reviewed the revised PWROG plan and fil")ds that the proposed alternatives match the inspection plan for the PWR fleet. . Based on the above evaluation, the NRC staff concludes that the licensee has addressed specific information item 3 (see SE Section 2.4) satisfactorily because the licensee demonstrated that the plant-specific flaw information for WCGS in 13R-09 is bounded by the WCAP-A, supporting the plant-specific applicability of the WCAP-A to the WCGS. In summary, the NRC staff has reviewed the .licensee's submittal and the response to the NRC staff's RAI supplementing the relief request. The licensee has met all of the plant-specific requirements for application of the WCAP-A. In addition, the NRC staff found good agreement between its independent calculations compared to those from the licensee in Table 3 of the relief request. With this information, the NRC staff concluded that the TWCFss-TOTAL value in Table 3 of Attachment 2 to the January 8, 2014, submittal is bounded by the WCAP-A results. Consequently, the licensee has demonstrated that the proposed alternative will provide an acceptable level of quality and safety and meets the guidance provided by RG 1.174, Rev. 1 for risk-informed decisions.

4.0 CONCLUSION

The NRC staff has completed its review of RRs 13R-08 and 13R-09 for WCGS. For the later, the staff concludes that increasing the lSI interval for the UT examination of Categories 8-A and 8-D components from 10 to 20 years will result in no appreciable increase in risk. This conclusion is based on the fact that the plant-specific information provided by the licensee is bounded by the data in the WCAP-A, and that the request meets all conditions and limitations in the NRC approval of the WCAP-A methodology.

Therefore, RR 13R-09 provides an acceptable level of quality and safety, and the alternative is authorized for Categories 8-A and 8-D components pursuant to 10 CFR 50.55a(a)(3)(i) until the end of the third 1 0-year lSI interval, which is now 2025 for WCGS. Further, the NRC staff accepts the alternative examination date of 2025 for Categories 8-A and 8-D components for the WCGS RPV. Furthermore, given the approval of RR 13R-09, the NRC staff concludes that deferring the lSI VT-3 examination of Categories 8-N-2 and 8-N-3 components from RF20 to RF21, 3 months beyond the maximum lSI interval length allowed by the ASME Code, is acceptable because it minimizes the risk associated with removal of the core barrel and fuel and follows the ALARA principles.

Requiring the licensee to follow the Code requirements would represent a hardship without a compensating increase in the level of quality and safety: Therefore, RR 13R-08 is authorized for Categories B-N-2 and B-N-3 components pursuant to 10 CFR 50.55a(a)(3)(ii) until the end of RF21, currently scheduled for the fall of 2016. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor:

P. Purtscher, NRR/DE/EVIB Date: December 10, 2014 .

A. Heflin The detailed results of the NRC staff review are provided in the enclosed safety evaluation.

If you have any questions concerning this matter, please call Mr. F. Lyon of my staff at (301) 415-2296 or by electronic mail at fred.lyon@nrc.gov.

Docket No. 50-482 Enclosure Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL4-1 r/f RidsAcrsAcnw'-MaiiCTR Resource RidsNrrDeEvib Resource RidsNrrDorllpl4-1 Resource ADAMS Accession No* ML 14321A864

.. Sincerely, IRA/ Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrLAJBurkhardt Resource RidsNrrPMWolfCreek Resource RidsRgn4MaiiCenter Resource PPurtscher, NRR/DE/EVIB TSantos, EDO RIV *email dated OFFICE 1/PM NRR/DORLILPL4-1/LA NRR/DE/EVIB/BC NRR/DORL/LPL4-1/BC(A)

NAME Flyon JBurkhardt SRosenberg*

EOesterle DATE 12/9/14 12/9/14 11/12/14 12/10/14 OFFICIAL RECORD COPY