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| | number = ML14363A409 | | | number = ML14363A409 |
| | issue date = 01/28/2015 | | | issue date = 01/28/2015 |
| | title = Beaver Valley Power Station, Unit 1, Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1 (TAC No. MF5049) | | | title = Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1 |
| | author name = Khanna M K | | | author name = Khanna M |
| | author affiliation = NRC/NRR/DORL/LPLI-2 | | | author affiliation = NRC/NRR/DORL/LPLI-2 |
| | addressee name = Larson E A | | | addressee name = Larson E |
| | addressee affiliation = FirstEnergy Nuclear Operating Co | | | addressee affiliation = FirstEnergy Nuclear Operating Co |
| | docket = 05000334, 05000412 | | | docket = 05000334, 05000412 |
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| | page count = 9 | | | page count = 9 |
| | project = TAC:MF5049 | | | project = TAC:MF5049 |
| | | stage = Other |
| }} | | }} |
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| =Text= | | =Text= |
| {{#Wiki_filter: | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 28, 2015 Mr. Eric A. Larson, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077 |
| [[Issue date::January 28, 2015]]
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| Mr. Eric A. Larson, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077
| | ==SUBJECT:== |
| | BEAVER VALLEY POWER STATION, UNIT NO.1- RELIEF REQUEST NO. 1TYP-4-RV-04 REGARDING THE EXAMINATION REQUIREMENTS OF CODE CASE N-729-1 (TAC NO. MF5049) |
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| SUBJECT: BEAVER VALLEY POWER STATION, UNIT NO.1-RELIEF REQUEST NO. 1TYP-4-RV-04 REGARDING THE EXAMINATION REQUIREMENTS OF CODE CASE N-729-1 (TAC NO. MF5049)
| | ==Dear Mr. Larson:== |
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| ==Dear Mr. Larson:==
| | By letter dated October 17, 2014, 1 FirstEnergy Nuclear Operating Company (the licensee) submitted request 1TYP-4-RV-04 to the Nuclear Regulatory Commission (NRC). The licensee requested to use alternative requirements to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, "Rules for lnservice Inspection (lSI) of Nuclear Power Plant Components," for Beaver Valley Power Station, Unit 1 (BVPS-1). |
| By letter dated October 17, 2014,1 FirstEnergy Nuclear Operating Company (the licensee) submitted request 1TYP-4-RV-04 to the Nuclear Regulatory Commission (NRC). The licensee requested to use alternative requirements to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, "Rules for lnservice Inspection (lSI) of Nuclear Power Plant Components," for Beaver Valley Power Station, Unit 1 (BVPS-1). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.55a(a)(3)(i), (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested an alternative to performing the required volumetric and surface examinations for certain reactor pressure head components at the frequency prescribed in ASME Code Case N-729-1. The NRC staff has reviewed the licensee's relief request and has determined that the requested alternative will provide an acceptable level of quality and safety, as documented in the enclosed safety evaluation. Therefore, the licensee's request for the use of the above stated alternative is authorized pursuant to 10 CFR 50.55a(z)(1) for the BVPS-1. 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 lnservice Inspector. 1 Agencywide Documents Access and Management System Accession No. ML 14290A140. If you have any questions, please contact the Beaver Valley Project Manager, Taylor A. Lamb, at (301) 415-7128 or Taylor.Lamb@ nrc.gov. Docket No. 50-334 | | Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) |
| | Section 50.55a(a)(3)(i), (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested an alternative to performing the required volumetric and surface examinations for certain reactor pressure head components at the frequency prescribed in ASME Code Case N-729-1. |
| | The NRC staff has reviewed the licensee's relief request and has determined that the requested alternative will provide an acceptable level of quality and safety, as documented in the enclosed safety evaluation. Therefore, the licensee's request for the use of the above stated alternative is authorized pursuant to 10 CFR 50.55a(z)(1) for the BVPS-1. |
| | 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 lnservice Inspector. |
| | 1 Agencywide Documents Access and Management System Accession No. ML14290A140. |
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| | E. Larson If you have any questions, please contact the Beaver Valley Project Manager, Taylor A. Lamb, at (301) 415-7128 or Taylor.Lamb@ nrc.gov. |
| | Sincerely, Meena Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-334 |
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| | ==Enclosure:== |
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| | Safety Evaluation cc w/encl: Distribution via Listserv |
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| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST REGARDING THE PROPOSED ALTERNATIVE TO ASME CODE CASE N-729-1 EXAMINATION FREQUENCY REQUIREMENTS FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION. UNIT 1 DOCKET NUMBER 50-334 |
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| | ==1.0 INTRODUCTION== |
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| | By letter dated October 17, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML14290A140), FirstEnergy Nuclear Operating Company (the licensee or FENOC) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, associated with the examination frequency requirements of Code Case N-729-1 for Beaver Valley Power Station, Unit No. 1 (BVPS-1 ). |
| | Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) |
| | Section 50.55a(a)(3)(i) (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested to use the proposed alternatives in Relief Request 1TYP-4-RV-04, to the examination frequency of ASME Code Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1,"on the basis that the alternative examination provides an acceptable level of quality and safety. |
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| | ==2.0 REGULATORY EVALUATION== |
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| | The regulations in 10 CFR 50.55a(g)(6)(ii) state that "the Commission may require the licensee to follow an augmented inservice inspection [lSI] program for systems and components for which the Commission deems that added assurance of structural reliability is necessary." |
| | The regulations in 10 CFR 50.55a(g)(6)(ii)(D), state, in part, that "all licensees of pressurized water reactors shall augment their [lSI] program with ASME Code Case N 729-1, subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section ... " |
| | In this request, the licensee has requested relief from the examination frequency required by Code Case N-729-1 and has, therefore, also requested relief from 10 CFR 50.55a(g)(6)(ii)(D). |
| | Enclosure |
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| | The regulations in 10 CFR 50.55a(z) state that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that (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. |
| | 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 Commission to authorize the proposed alternative requested by the licensee. |
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| | ==3.0 TECHNICAL EVALUATION== |
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| | 3.1 COMPONENTS AFFECTED The affected components are ASME Class 1, Reactor Vessel Closure Head (RVCH) |
| | Penetration Nozzle numbers 1 through 62, which are fabricated from lnconel SB-167 (Alloy 690) |
| | UNS N06690. The nozzle J-groove welds are fabricated from UNS N06052 and UNS W86152, 52/152 weld materials. The original BVPS-1 RVCH penetration nozzles, which were manufactured with Alloy 600/82/182 materials, was replaced with a new RVCH using Alloy 690/52/152 material for the penetration nozzles during the refueling outage that returned to operation in April 2006. |
| | 3.2 INSERVICE INSPECTION INTERVAL The licensee's current lSI is the fourth 10-year lSI, which started on April 1, 2008, and ends on March 31, 2018. The proposed duration of the alternative is requested for the duration up to and including the 251h BVPS-1 refueling outage that is scheduled to commence in April2018. |
| | 3.3 ASME CODE OF RECORD The ASME Section XI Code of Record for the current, fourth 10-year lSI interval at BVPS-1, is the 2001 Edition through the 2003 Addenda. |
| | 3.4 ASME CODE AND/OR REGULATORY REQUIREMENTS Section 50.55a(g)(6)(ii)(D) of 10 CFR requires, in part, that licensees augment their lSI program in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D). ASME Code Case N-729-1, Table 1, Inspection Item 84.40 requires volumetric/surface examination be performed within one inspection interval (nominally 10 calendar years) of its in service date for a replaced RVCH. The required volumetric/surface examinations would thus have to be completed by early 2016 in order to fulfill the requirements of N-729-1. |
| | 3.5 PROPOSED ALTERNATIVE The licensee proposes to delay the next required inspection for a period of approximately 2 years. The licensee proposes to accomplish the inspection in accordance with ASME Code |
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| | Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D) during refueling outage 25, which is scheduled for April2018. |
| | 3.6 LICENSEE'S BASIS FOR USE OF THE PROPOSED ALTERNATIVE The licensee's basis for use of the proposed alternative is based primarily on three topics of consideration. The first topic addresses the concept that the inspection interval in Code Case N-729-1 is based on primary water stress-corrosion cracking (PWSCC) crack growth rates for Alloy 600/82/182. The second topic addresses a bare metal visual examination that was conducted on the licensee's replacement RVCH in 2010. The third topic addresses a plant-specific factor of improvement analysis that was conducted by the licensee. |
| | In addressing its first basis for use of the proposed alternative, the licensee asserts that the inspection intervals contained in ASME Code Case N-729-1 for alloy 600/82/182 are based on re-inspection years ~RIY) equal to 2.25. This RIY value is based on PWSCC crack growth rates as defined in the 751 percentile curve contained in Materials Reliability Program (MRP) 55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," and MRP 115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds," both NRC approved documents. The licensee further asserts that the PWSCC crack growth rates of Alloy 690/52/152 are significantly lower than those of Alloy 600/82/182 and, therefore, merit a longer inspection interval. The licensee bases that assertion on: (1) the lack of cracking in other Alloy 690 components, such as steam generators and pressurizers, in the nearly 20 years that Alloy 690 has been in service in these components; (ii) the failure to observe cracking in inspections already performed in replacement heads (13 of 40 replacement heads in the United States have been examined, which includes heads that operate at higher temperatures than the head under consideration); (iii) the similarity of the inspected heads to the head under consideration regarding configuration, manufacturers, design and operating conditions; and (iv) laboratory test data for Alloy 690/52/152 as contained in MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles." |
| | In addressing its second basis for use of the proposed alternative, the licensee stated that a bare metal visual examination was performed in 2010 on the BVPS-1 replacement RVCH in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This visual examination was performed by VT -2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any indications of nozzle leakage (e.g., boric acid deposits) on the surface or near a nozzle penetration. The licensee also indicated that this examination will be performed again in the upcoming refueling outage, which is scheduled to commence in April 2015. Also, the licensee stated that no alternative examination processes are proposed to those required by ASME Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The visual (VT-2) examinations and acceptance criteria, as required by Item B4.30 of Table 1 of ASME Code Case N-729-1, are not affected by this request and will continue to be performed on a frequency not to exceed every 5 calendar years. |
| | In addressing its third basis for use of the proposed alternative, the licensee made a plant-specific calculation of the required factor of improvement in the crack growth rate of Alloy |
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| | 690/52/152, as compared to the crack growth rate of Alloy 600/82/182. In making this calculation, the licensee used the actual temperature of the head and conservatively assumed that calendar years were equal to effective full-power years. Based on this calculation, the licensee determined that an improvement factor of 5.5 was required to meet the proposed and desired inspection interval of 12 calendar years. The licensee then proposed that, because the required factor of improvement (5.5) was smaller than the factor of improvement of 20, which bounded most of the MRP 375 data for Alloy 690/52/152, the use of a factor of improvement of 5.5 would not result in a reduction in safety and was, therefore, justified. |
| | The licensee stated that their analysis showed significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the BVPS-1 replacement RVCH provide for a reactor coolant system pressure boundary, where the potential for PWSCC has been shown by analysis and by years of positive industry experience, to be remote. As such, the licensee found the technical basis sufficient to ensure public health and safety by extending the inspection frequency of the RVCH nozzle at BVPS-1 from a maximum of 10 years to a new maximum of 12 years. |
| | 3.7 NRC STAFF EVALUATION In evaluating the technical sufficiency of the licensee's proposed alternative (i.e., a one-time extension of the volumetric/surface examination interval contained in ASME Code Case N-729-1 from 10 years to no longer than 12 years), the NRC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative. The NRC staff found that the technical basis included by the licensee provided sufficient information for the NRC staff to review the proposed alternative. |
| | Due to concerns about PWSCC, many PWR plants in the United States and overseas have replaced reactor vessel closure heads containing Alloy 600/182/82 nozzles with heads containing Alloy 690/152/52 nozzles. The inspection frequencies developed in Code Case N-729-1 for RVCH penetration nozzles using Alloy 600/182/82 were developed based, in part, on those material's crack growth rate equations, as documented in MRP 55 and MRP 115. The licensee's primary technical basis is to present crack growth rate data for the new, more crack resistant materials, Alloy 690/152/52, and demonstrate an improvement factor (IF) of these materials versus the older Alloy 600/82/182 materials. This IF would then provide the basis for the extension of the lSI frequency requested by the licensee in their proposed alternative. |
| | In evaluating the licensee's first technical basis for use of the proposed alternative, the NRC staff notes that the licensee used MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles." This document, in part, summarizes numerous Alloy 690/152/52 crack growth rate data from various sources to develop IFs for the crack growth rate equations provided in MRP 55 and MRP 115. While the NRC staff finds the licensee's assertions and/or interpretations to be reasonable, MRP 375 is not an NRC approved document. As the NRC staff does not have sufficient time or resources to validate all of the data used by this document, the NRC staff does not consider it appropriate to use all of the data from this document to review the licensee's relief request. A more detailed review of the data provided in MRP 375 will be performed by an international group of experts as part of an Alloy 690 Expert Panel, which is currently scheduled to complete its review in the |
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| | 2016-2017 timeframe. In the interim, the NRC staff review will rely upon Alloy 690/152/52 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data is documented in a data summary report and car:~ be found under ADAMS Accession Number ML14322A587. The NRC confirmatory research generally supports the contention that the crack growth rate of Alloy 690/52/152 is more crack resistant but differs from the MRP 375 data in some respects. |
| | The PNNL and ANL data summary report includes crack growth rate data up to approximately 20 percent cold work based on the observation of local strains in welds and weld dilution zone data. However, the NRC staff did not consider the weld dilution zone data in its assessment. |
| | This is because the limited weld dilution zone data that is currently available has shown higher crack growth rates than are commonly observed for Alloy 690/152/52 material. The high crack growth rates in weld dilution zones may be due to the reduced chromium present in these areas. The NRC staff chose to exclude the weld dilution zone data from this analysis due to the limited number of data points available, the variability in results, and due to the limited area of continuous weld dilution for flaws to grow through. For example, in the case of the highest measured crack growth rates, a flaw would have to travel in the heat affected zone of a j-groove weld along the low Alloy steel head interface. It is not fully apparent to the NRC staff how accelerated crack growth in very small areas of the weld dilution zone would result in a significantly increased probability of leakage or component failure during a relatively short extension of the required inspection interval. Exclusion of these data may be reevaluated as additional data become available; a better understanding of the existing data is obtained; or if a longer extension of the inspection interval is requested. Therefore, the NRC staff finds that the impact of these weld dilution zone crack growth rates on the change in volumetric inspection frequency, as requested by the licensee's proposed alternative, is not considered to be relevant for this specific relief request. |
| | In evaluating the licensee's second basis for use of the proposed alternative, the NRC staff finds that the past bare metal visual examination on the head under consideration is a reasonable means to demonstrate the absence of leakage through the nozzle/J-groove weld, prior to the time the examination was conducted. The NRC staff also finds that performance of future bare metal visual examinations, in accordance with the code case, is adequate to demonstrate the absence of leakage at or prior to the time the examinations are conducted. Finally, the NRC staff finds that the proposed alternative's frequency for bare metal visual examinations, in conjunction with the new frequency of volumetric examinations, is sufficient to provide reasonable assurance of the structural integrity of the RVCH. |
| | In evaluating the licensee's third basis for use of the proposed alternative, the NRC staff found that the licensee's calculated improvement factor of 5.5, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 12 calendar years, was found to be acceptable by NRC staff calculation. The NRC staff also found that the application of an IF of 5.5 to the 75th percentile curves in MRP 55 and 115 bounded essentially all of the NRC data included in the PNNL and ANL data summary report. Therefore, the NRC staff found that this analysis supports the concept that a volumetric inspection interval for the RVCH of not more than 12 calendar years does not pose a higher risk than that associated with an Alloy 600/182/82 RVCH inspected at intervals of 2.25 RIY. Hence, the NRC staff found the licensee's technical basis to be acceptable. |
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| | Therefore, based on the above evaluation, the NRC finds that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) |
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| ===Enclosure:=== | | ==4.0 CONCLUSION== |
| Safety Evaluation cc w/encl: Distribution via Listserv
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| Sincerely,Meena Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST REGARDING THE PROPOSED ALTERNATIVE TO ASME CODE CASE N-729-1 EXAMINATION FREQUENCY REQUIREMENTS FIRSTENERGY NUCLEAR OPERATING COMPANY 1.0 INTRODUCTION BEAVER VALLEY POWER STATION. UNIT 1 DOCKET NUMBER 50-334 By letter dated October 17, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML 14290A140), FirstEnergy Nuclear Operating Company (the licensee or FENOC) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, associated with the examination frequency requirements of Code Case N-729-1 for Beaver Valley Power Station, Unit No. 1 (BVPS-1 ). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.55a(a)(3)(i) (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested to use the proposed alternatives in Relief Request 1TYP-4-RV-04, to the examination frequency of ASME Code Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1 ,"on the basis that the alternative examination provides an acceptable level of quality and safety. 2.0 REGULATORY EVALUATION The regulations in 10 CFR 50.55a(g)(6)(ii) state that "the Commission may require the licensee to follow an augmented inservice inspection [lSI] program for systems and components for which the Commission deems that added assurance of structural reliability is necessary." The regulations in 10 CFR 50.55a(g)(6)(ii)(D), state, in part, that "all licensees of pressurized water reactors shall augment their [lSI] program with ASME Code Case N 729-1, subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section ... " In this request, the licensee has requested relief from the examination frequency required by Code Case N-729-1 and has, therefore, also requested relief from 10 CFR 50.55a(g)(6)(ii)(D). Enclosure
| | As set forth above, the NRC staff has determined that the alternative method proposed by the licensee in 1TYP-4-RV-04, Revision 0, will provide an acceptable level of quality and safety for the examination frequency requirements of the RVCH. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the one time use of 1TYP-4-RV-04 at Beaver Valley Power Station, Unit No.1, for the duration up to and including the 25th refueling outage that is scheduled to commence in April 2018. |
| -2-The regulations in 10 CFR 50.55a(z) state that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that (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. 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 Commission to authorize the proposed alternative requested by the licensee. 3.0 TECHNICAL EVALUATION 3.1 COMPONENTS AFFECTED The affected components are ASME Class 1, Reactor Vessel Closure Head (RVCH) Penetration Nozzle numbers 1 through 62, which are fabricated from lnconel SB-167 (Alloy 690) UNS N06690. The nozzle J-groove welds are fabricated from UNS N06052 and UNS W86152, 52/152 weld materials. The original BVPS-1 RVCH penetration nozzles, which were manufactured with Alloy 600/82/182 materials, was replaced with a new RVCH using Alloy 690/52/152 material for the penetration nozzles during the refueling outage that returned to operation in April 2006. 3.2 INSERVICE INSPECTION INTERVAL The licensee's current lSI is the fourth 1 0-year lSI, which started on April 1, 2008, and ends on March 31, 2018. The proposed duration of the alternative is requested for the duration up to and including the 251h BVPS-1 refueling outage that is scheduled to commence in April2018. 3.3 ASME CODE OF RECORD The ASME Section XI Code of Record for the current, fourth 1 0-year lSI interval at BVPS-1, is the 2001 Edition through the 2003 Addenda. 3.4 ASME CODE AND/OR REGULATORY REQUIREMENTS Section 50.55a(g)(6)(ii)(D) of 1 0 CFR requires, in part, that licensees augment their lSI program in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D). ASME Code Case N-729-1, Table 1, Inspection Item 84.40 requires volumetric/surface examination be performed within one inspection interval (nominally 10 calendar years) of its in service date for a replaced RVCH. The required volumetric/surface examinations would thus have to be completed by early 2016 in order to fulfill the requirements of N-729-1. 3.5 PROPOSED ALTERNATIVE The licensee proposes to delay the next required inspection for a period of approximately 2 years. The licensee proposes to accomplish the inspection in accordance with ASME Code
| | All other requirements of the ASME Code, Section XI, and 10 CFR 50.55a(g)(6)(ii)(D) 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. |
| -3-Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D) during refueling outage 25, which is scheduled for April2018. 3.6 LICENSEE'S BASIS FOR USE OF THE PROPOSED ALTERNATIVE The licensee's basis for use of the proposed alternative is based primarily on three topics of consideration. The first topic addresses the concept that the inspection interval in Code Case N-729-1 is based on primary water stress-corrosion cracking (PWSCC) crack growth rates for Alloy 600/82/182. The second topic addresses a bare metal visual examination that was conducted on the licensee's replacement RVCH in 2010. The third topic addresses a specific factor of improvement analysis that was conducted by the licensee. In addressing its first basis for use of the proposed alternative, the licensee asserts that the inspection intervals contained in ASME Code Case N-729-1 for alloy 600/82/182 are based on re-inspection years equal to 2.25. This RIY value is based on PWSCC crack growth rates as defined in the 751 percentile curve contained in Materials Reliability Program (MRP) 55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," and MRP 115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds," both NRC approved documents. The licensee further asserts that the PWSCC crack growth rates of Alloy 690/52/152 are significantly lower than those of Alloy 600/82/182 and, therefore, merit a longer inspection interval. The licensee bases that assertion on: (1) the lack of cracking in other Alloy 690 components, such as steam generators and pressurizers, in the nearly 20 years that Alloy 690 has been in service in these components; (ii) the failure to observe cracking in inspections already performed in replacement heads (13 of 40 replacement heads in the United States have been examined, which includes heads that operate at higher temperatures than the head under consideration); (iii) the similarity of the inspected heads to the head under consideration regarding configuration, manufacturers, design and operating conditions; and (iv) laboratory test data for Alloy 690/52/152 as contained in MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles." In addressing its second basis for use of the proposed alternative, the licensee stated that a bare metal visual examination was performed in 2010 on the BVPS-1 replacement RVCH in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This visual examination was performed by VT -2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any indications of nozzle leakage (e.g., boric acid deposits) on the surface or near a nozzle penetration. The licensee also indicated that this examination will be performed again in the upcoming refueling outage, which is scheduled to commence in April 2015. Also, the licensee stated that no alternative examination processes are proposed to those required by ASME Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The visual (VT-2) examinations and acceptance criteria, as required by Item B4.30 of Table 1 of ASME Code Case N-729-1, are not affected by this request and will continue to be performed on a frequency not to exceed every 5 calendar years. In addressing its third basis for use of the proposed alternative, the licensee made a specific calculation of the required factor of improvement in the crack growth rate of Alloy
| | Principal Contributor: Steven Vitto Date: January 28, 2015 |
| -4-690/52/152, as compared to the crack growth rate of Alloy 600/82/182. In making this calculation, the licensee used the actual temperature of the head and conservatively assumed that calendar years were equal to effective full-power years. Based on this calculation, the licensee determined that an improvement factor of 5.5 was required to meet the proposed and desired inspection interval of 12 calendar years. The licensee then proposed that, because the required factor of improvement (5.5) was smaller than the factor of improvement of 20, which bounded most of the MRP 375 data for Alloy 690/52/152, the use of a factor of improvement of 5.5 would not result in a reduction in safety and was, therefore, justified. The licensee stated that their analysis showed significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the BVPS-1 replacement RVCH provide for a reactor coolant system pressure boundary, where the potential for PWSCC has been shown by analysis and by years of positive industry experience, to be remote. As such, the licensee found the technical basis sufficient to ensure public health and safety by extending the inspection frequency of the RVCH nozzle at BVPS-1 from a maximum of 10 years to a new maximum of 12 years. 3.7 NRC STAFF EVALUATION In evaluating the technical sufficiency of the licensee's proposed alternative (i.e., a one-time extension of the volumetric/surface examination interval contained in ASME Code Case N-729-1 from 10 years to no longer than 12 years), the NRC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative. The NRC staff found that the technical basis included by the licensee provided sufficient information for the NRC staff to review the proposed alternative. Due to concerns about PWSCC, many PWR plants in the United States and overseas have replaced reactor vessel closure heads containing Alloy 600/182/82 nozzles with heads containing Alloy 690/152/52 nozzles. The inspection frequencies developed in Code Case N-729-1 for RVCH penetration nozzles using Alloy 600/182/82 were developed based, in part, on those material's crack growth rate equations, as documented in MRP 55 and MRP 115. The licensee's primary technical basis is to present crack growth rate data for the new, more crack resistant materials, Alloy 690/152/52, and demonstrate an improvement factor (IF) of these materials versus the older Alloy 600/82/182 materials. This IF would then provide the basis for the extension of the lSI frequency requested by the licensee in their proposed alternative. In evaluating the licensee's first technical basis for use of the proposed alternative, the NRC staff notes that the licensee used MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles." This document, in part, summarizes numerous Alloy 690/152/52 crack growth rate data from various sources to develop I Fs for the crack growth rate equations provided in MRP 55 and MRP 115. While the NRC staff finds the licensee's assertions and/or interpretations to be reasonable, MRP 375 is not an NRC approved document. As the NRC staff does not have sufficient time or resources to validate all of the data used by this document, the NRC staff does not consider it appropriate to use all of the data from this document to review the licensee's relief request. A more detailed review of the data provided in MRP 375 will be performed by an international group of experts as part of an Alloy 690 Expert Panel, which is currently scheduled to complete its review in the
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| -5-2016-2017 timeframe. In the interim, the NRC staff review will rely upon Alloy 690/152/52 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data is documented in a data summary report and be found under ADAMS Accession Number ML 14322A587. The NRC confirmatory research generally supports the contention that the crack growth rate of Alloy 690/52/152 is more crack resistant but differs from the MRP 375 data in some respects. The PNNL and ANL data summary report includes crack growth rate data up to approximately 20 percent cold work based on the observation of local strains in welds and weld dilution zone data. However, the NRC staff did not consider the weld dilution zone data in its assessment. This is because the limited weld dilution zone data that is currently available has shown higher crack growth rates than are commonly observed for Alloy 690/152/52 material. The high crack growth rates in weld dilution zones may be due to the reduced chromium present in these areas. The NRC staff chose to exclude the weld dilution zone data from this analysis due to the limited number of data points available, the variability in results, and due to the limited area of continuous weld dilution for flaws to grow through. For example, in the case of the highest measured crack growth rates, a flaw would have to travel in the heat affected zone of a j-groove weld along the low Alloy steel head interface. It is not fully apparent to the NRC staff how accelerated crack growth in very small areas of the weld dilution zone would result in a significantly increased probability of leakage or component failure during a relatively short extension of the required inspection interval. Exclusion of these data may be reevaluated as additional data become available; a better understanding of the existing data is obtained; or if a longer extension of the inspection interval is requested. Therefore, the NRC staff finds that the impact of these weld dilution zone crack growth rates on the change in volumetric inspection frequency, as requested by the licensee's proposed alternative, is not considered to be relevant for this specific relief request. In evaluating the licensee's second basis for use of the proposed alternative, the NRC staff finds that the past bare metal visual examination on the head under consideration is a reasonable means to demonstrate the absence of leakage through the nozzle/J-groove weld, prior to the time the examination was conducted. The NRC staff also finds that performance of future bare metal visual examinations, in accordance with the code case, is adequate to demonstrate the absence of leakage at or prior to the time the examinations are conducted. Finally, the NRC staff finds that the proposed alternative's frequency for bare metal visual examinations, in conjunction with the new frequency of volumetric examinations, is sufficient to provide reasonable assurance of the structural integrity of the RVCH. In evaluating the licensee's third basis for use of the proposed alternative, the NRC staff found that the licensee's calculated improvement factor of 5.5, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 12 calendar years, was found to be acceptable by NRC staff calculation. The NRC staff also found that the application of an IF of 5.5 to the 75th percentile curves in MRP 55 and 115 bounded essentially all of the NRC data included in the PNNL and ANL data summary report. Therefore, the NRC staff found that this analysis supports the concept that a volumetric inspection interval for the RVCH of not more than 12 calendar years does not pose a higher risk than that associated with an Alloy 600/182/82 RVCH inspected at intervals of 2.25 RIY. Hence, the NRC staff found the licensee's technical basis to be acceptable.
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| -6-Therefore, based on the above evaluation, the NRC finds that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) 4.0 CONCLUSION As set forth above, the NRC staff has determined that the alternative method proposed by the licensee in 1 TYP-4-RV-04, Revision 0, will provide an acceptable level of quality and safety for the examination frequency requirements of the RVCH. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the one time use of 1TYP-4-RV-04 at Beaver Valley Power Station, Unit No.1, for the duration up to and including the 25th refueling outage that is scheduled to commence in April 2018. All other requirements of the ASME Code, Section XI, and 10 CFR 50.55a(g)(6)(ii)(D) 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 Contributor: Steven Vitto Date: January 28, 2015}} | | ML14363A409 *via e-mail OFFICE LPL 1-2/PM LPL 1-2/PM LPL 1-1/LA* DE/EMCB/BC* LPL 1-1/BC LPL 1-2/PM NAME TLamb JWhited A Baxter DAIIey MKhanna TLamb DATE 1/20/15 1/20/15 1/28/15 12/19/14 1/21/15 1/28/15}} |
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MONTHYEARML14310A8532014-11-0606 November 2014 E-mail from J. Whited to P. Lashley Beaver Valley Power Station Unit 1 Acceptance of Requested Licensing Action Relief Request 1TYP-4-RV-04 Project stage: Acceptance Review ML14363A4092015-01-28028 January 2015 Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1 Project stage: Other 2014-11-06
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Category:Code Relief or Alternative
MONTHYEARML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 ML22082A2532022-03-29029 March 2022 Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27 ML22012A2972022-01-21021 January 2022 Proposed Alternative to Use ASME OM Code Case OMN-27 L-20-256, Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2020-09-28028 September 2020 Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI ML20223A0142020-08-20020 August 2020 Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19)) ML20206K8652020-08-0707 August 2020 Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19)) ML20145A0002020-06-0303 June 2020 Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19)) ML20100N3222020-04-0909 April 2020 Verbal Relief for Penetration Evaluation and Hot Leg Nozzles - Delivered 4/9/2020 at 10:00 Am ML20099B2572020-04-0808 April 2020 Verbal Relief Unit 2 CIV ML20098F3012020-04-0707 April 2020 Verbal Relief for Appendix I Safety Relief Valves ML20095J2192020-04-0404 April 2020 Email Beaver Valley Power Station, Unit 2 - Verbal Relief for MOVs - Delivered 4/4/2020 at 4:00 Pm ML20095J0992020-04-0404 April 2020 Email Beaver Valley Power Station, Unit 2 - Verbal Relief for Snubbers - Delivered 4/4/2020 at 4:00 P.M L-20-117, 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections2020-04-0303 April 2020 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections L-20-118, CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing2020-04-0303 April 2020 CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing L-20-060, CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency2020-04-0202 April 2020 CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency L-20-116, CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency2020-04-0101 April 2020 CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency ML20079F8162020-03-26026 March 2020 Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval ML19275E2942019-10-16016 October 2019 Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements L-19-107, Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements2019-08-27027 August 2019 Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements ML19051A1082019-02-20020 February 2019 Request Alternative Examination Frequency for Reactor Vessel Nozzle to Safe-End Welds (Request 2-TYP-4-RV-05, Revision 0) ML18227A7332018-08-27027 August 2018 Request for Relief from the Requirements of the ASME Code ML18004A1222018-01-22022 January 2018 FENOC-Beaver Valley, Davis-Besse, and Perry - Alternative for the Use of ASME Code Case N-513-4 (CAC Nos. MG0120, MG0121, MG0122, and MG0123; EPID L-2017-LLR-0088) L-17-317, Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01)2017-11-15015 November 2017 Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01) L-17-308, 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)2017-10-25025 October 2017 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2) ML17167A0672017-06-26026 June 2017 Requests for Alternatives and Requests for Relief Fourth 10-Year Inservice Test Program Interval (CAC Nos. MF8333-MF8356). Note: Correction Safety Evaluation See ML17255A526 ML17159A4422017-06-26026 June 2017 Requests for Alternatives and Requests for Relief Fifth 10-Year Inservice Testing Program Interval (CAC Nos. MF8332 Through MF8357). Note: Correction Safety Evaluation See ML17255A508 ML17047A6102017-03-0202 March 2017 Relief from the Requirements of the American Society of Mechanical Engineers Code ML17041A1852017-03-0202 March 2017 Relief from the Requirements of the ASME Code ML17048A0042017-03-0202 March 2017 Relief from the Requirements of the ASME Code ML16328A1252017-01-23023 January 2017 Relief from the Requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) (TAC No. MF7780 - MF7783) ML16190A1332016-12-27027 December 2016 Relief from the Requirements of the ASME Code ML16319A0572016-12-0101 December 2016 Acceptance of Requested Licensing Action Relief Request for Proposed Alternative to Perform As-Found Set-Pressure Test ML16257A6212016-11-21021 November 2016 Relief Request BV2-PZR-01, Regarding Alternative to Requirements for Components Connected to the Steam Side of the Pressurizer ML16228A4082016-10-21021 October 2016 Correction to Relief Request 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds ML16147A3622016-06-17017 June 2016 Relief Request No. 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds ML14363A4092015-01-28028 January 2015 Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1 ML1202702982012-02-0707 February 2012 Relief Request VRR3 Regarding Solenoid Operated Valve Remote Position Verification Frequency ML1131304282011-11-22022 November 2011 Relief Request VRR5 Regarding Turbine Driven Auxiliary Feedwater Valve Test Frequency for the 10-Year Inservice Testing Program Interval ML1126404122011-09-20020 September 2011 Acceptance Review Results for VC Summer Relief Request (ME6879) ML1107705512011-04-26026 April 2011 Relief Request VRR2 Regarding the 10-Year Inservice Testing Program Interval ML1104705572011-02-25025 February 2011 Relief Request Regarding an Alternative Weld Repair Method for Reactor Vessel Head Penetrations J-Groove Welds ML1006807812010-03-12012 March 2010 Third 10-Year ISI Interval Relief Request (ME2608) L-08-362, Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements2008-12-0202 December 2008 Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements L-08-207, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement2008-09-24024 September 2008 Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement L-08-069, Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3)2008-04-0909 April 2008 Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3) ML0720504882007-09-17017 September 2007 Relief Request No. BV1-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds ML0705905552007-04-30030 April 2007 Relief, Relief Request No. BV2-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds L-07-056, Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update2007-03-28028 March 2007 Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update ML0625801202006-10-0202 October 2006 (BVPS-1 and 2), Inservice Inspection (ISI) Program, Alternative Examination of Reactor Coolant Pipe Welds, Request for Relief No. BV3-RV-2 L-06-042, Proposed Alternative to American Society of Mechanical Engineers Code Section XI Examination Requirements2006-04-0707 April 2006 Proposed Alternative to American Society of Mechanical Engineers Code Section XI Examination Requirements 2023-09-18
[Table view] Category:Letter
MONTHYEARML24025A0922024-01-25025 January 2024 Requalification Program Inspection L-23-267, Submittal of Discharge Monitoring Report (Npdes), Permit No. PA00256152023-12-18018 December 2023 Submittal of Discharge Monitoring Report (Npdes), Permit No. PA0025615 IR 05000334/20230112023-12-0404 December 2023 Age-Related Degradation Inspection Report 05000334/2023011 and 05000412/2023011 L-23-229, Request for Additional Information Regarding the Spring 2023 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F-Star Reports2023-11-29029 November 2023 Request for Additional Information Regarding the Spring 2023 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F-Star Reports IR 05000334/20233012023-11-27027 November 2023 Initial Operator Licensing Examination Report 05000334/2023301 L-23-247, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-11-17017 November 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 IR 05000334/20230032023-11-0606 November 2023 Integrated Inspection Report 05000334/2023003 and 05000412/2023003 L-23-227, Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Third Quarter 20232023-10-20020 October 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Third Quarter 2023 ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23198A3592023-10-0202 October 2023 Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary IR 05000334/20234022023-09-29029 September 2023 Security Baseline Inspection Report 05000334/2023402 and 05000412/2023402 (Cover Letter Only) ML23237B4222023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Letter Regarding Order Approving Transfer of Licenses and Draft Conforming License Amendments ML23269A1242023-09-27027 September 2023 Request for Withholding Information from Public Disclosure IR 05000334/20234012023-09-21021 September 2023 Cybersecurity Inspection Report 05000334/2023401 and 05000412/2023401 (Cover Letter Only) ML23263A0192023-09-20020 September 2023 Operator Licensing Examination Approval ML23243A9272023-09-19019 September 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 L-23-208, Submittal of Discharge Monitoring Report Cnpdes), Permit No. PA00256152023-09-14014 September 2023 Submittal of Discharge Monitoring Report Cnpdes), Permit No. PA0025615 L-23-167, Twenty-Third Refueling Outage Inservice Inspection Summary Report2023-09-13013 September 2023 Twenty-Third Refueling Outage Inservice Inspection Summary Report L-23-205, Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-09-12012 September 2023 Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments IR 05000334/20230052023-08-31031 August 2023 Updated Inspection Plan for Beaver Valley Power Station Units 1 and 2 (Report 05000334/2023005 and 05000412/2023005) L-23-172, Quality Assurance Program Manual2023-08-31031 August 2023 Quality Assurance Program Manual ML23129A1722023-08-25025 August 2023 Request for Withholding Information from Public Disclosure for Beaver Valley Power Station, Units 1 and 2; Davis Besse Nuclear Power Station, Unit 1; and Perry Nuclear Power Plant, Unit 1 IR 05000334/20230022023-08-0909 August 2023 Integrated Inspection Report 05000334/2023002 and 05000412/2023002 ML23219A0592023-08-0707 August 2023 Steam Generator Inspection Reports - Spring 2023 Spring Refueling Outrage L-23-188, Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-08-0707 August 2023 Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments L-23-179, Submittal of Discharge Monitoring Report, (NPDES) Permit No. PA00256152023-07-18018 July 2023 Submittal of Discharge Monitoring Report, (NPDES) Permit No. PA0025615 ML23188A0982023-07-17017 July 2023 Correction to the Safety Evaluation Issued Related to Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23188A0672023-07-10010 July 2023 Project Manager Assignment L-23-165, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-06-26026 June 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-23-139, Response to Request for Additional Information Regarding Fall 2022 180-Day Steam Generator Tube Inspection Report2023-06-13013 June 2023 Response to Request for Additional Information Regarding Fall 2022 180-Day Steam Generator Tube Inspection Report ML23157A1072023-06-0707 June 2023 Request for Information and Notification of Conduct of IP 71111.21.N.04, Age-Related Degradation, Reference Inspection Report 05000334/2023011 and 05000412/2023011 ML23143A0272023-05-23023 May 2023 Licensed Operator Positive Fitness-For-Duty Test L-23-055, Submittal of the Updated Final Safety Analysis Report, Revision 342023-05-23023 May 2023 Submittal of the Updated Final Safety Analysis Report, Revision 34 ML23102A1472023-05-22022 May 2023 Issuance of Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time L-23-065, Annual Financial Report2023-05-22022 May 2023 Annual Financial Report ML23135A7872023-05-19019 May 2023 Summary of Conference Calls Regarding the Spring 2023 Steam Generator Tube Inspections L-23-137, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-05-18018 May 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 ML23124A1742023-05-17017 May 2023 Energy Harbor Fleet Vistra License Transfer - Request for Withholding Information from Public Disclosure for Commance Peak Plant, Units 1 & 2, Beaver Valley Station, Units 1 & 2, Davis Besse Station, Unit 1 and Perry Plant, Unit 1 L-23-125, Cycle 24 Core Operating Limits Report2023-05-17017 May 2023 Cycle 24 Core Operating Limits Report ML23124A3882023-05-16016 May 2023 Summary of Regulatory Audit in Support of License Amendment Request to Revise Technical Specification 5.6.3, Core Operating Limits Reports (COLR) and TS 4.2.1 Fuel Assemblies ML23129A0112023-05-16016 May 2023 Notice of Consideration of Approval of Indirect and Direct License Transfer for Comanche Peak Plant, Units 1 & 2, Beaver Valley Station, Units 1 & 2, Davis Besse Station, Unit 1 and Perry Plant, Unit 1 (EPID L-2023-LLM-0000) (Letter) L-23-132, Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations2023-05-10010 May 2023 Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations IR 05000334/20230012023-05-0808 May 2023 Integrated Inspection Report 05000334/2023001 and 05000412/2023001 L-23-129, Response to Request for Additional Information for 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations2023-05-0505 May 2023 Response to Request for Additional Information for 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations ML23118A3812023-04-28028 April 2023 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations L-23-115, Submittal of 2022 Annual Radioactive Effluent Release Report, 2022 Annual Radiological Environmental Operating Report, and 2022 Annual Environmental Operating Report (Non-Radiological2023-04-27027 April 2023 Submittal of 2022 Annual Radioactive Effluent Release Report, 2022 Annual Radiological Environmental Operating Report, and 2022 Annual Environmental Operating Report (Non-Radiological L-23-126, Discharge Monitoring Report (Npdes), Permit No. PA00256152023-04-22022 April 2023 Discharge Monitoring Report (Npdes), Permit No. PA0025615 CP-202300181, ISFSI, Beaver Valley, Units 1 and 2, ISFSI, Davis-Besse, Unit 1, ISFSI, Perry, Unit 1, ISFSI, Corrected Affidavit for Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-04-20020 April 2023 ISFSI, Beaver Valley, Units 1 and 2, ISFSI, Davis-Besse, Unit 1, ISFSI, Perry, Unit 1, ISFSI, Corrected Affidavit for Application for Order Consenting to Transfer of Licenses and Conforming License Amendments ML23107A2502023-04-18018 April 2023 Requalification Program Inspection 2024-01-25
[Table view] Category:Safety Evaluation
MONTHYEARML23198A3592023-10-0202 October 2023 Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary ML23237B4282023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 2, Draft Conforming License Amendments ML23237B4302023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (EPID L-2023-LLM-0000) (Public) ML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 ML23188A0982023-07-17017 July 2023 Correction to the Safety Evaluation Issued Related to Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23102A1472023-05-22022 May 2023 Issuance of Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23019A0032023-03-16016 March 2023 Issuance of Amendment Nos. 320 and 210 Adoption of Technical Specifications Task Force (Tstf) Traveler Tstf-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML23062A5212023-03-0606 March 2023 Issuance of Amendment No. 319 Revise Technical Specification (TS) 3.5.2, ECCS Operating, Limiting Condition for Operation (LCO) 3.5.2, ML22277A8142022-10-0707 October 2022 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan ML22210A0102022-09-16016 September 2022 Energy Harbor Fleet- Issuance of Amendments Regarding Adoption of TSTF 554, Revise Reactor Coolant Leakage Requirements ML22222A0862022-09-0101 September 2022 Issuance of Amendment Nos. 317 and 208 Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start and Bus Separation Instrumentation ML22235A7672022-09-0101 September 2022 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan ML22202A4642022-06-29029 June 2022 Emergency Plan Safety Evaluation ML22140A2092022-06-28028 June 2022 Issuance of Amendment No. 207 Correct TS 3.1.7 Change Made by TSTF-547 ML22095A2352022-05-10010 May 2022 Issuance of Amendment Nos. 316 and 206 Revise Technical Specification 5.6.3, Core Operating Limits Report (COLR) ML21286A7822022-05-0606 May 2022 Issuance of Amendment Nos. 315 and 205 Regarding Changes to the Emergency Plan ML22077A1342022-05-0202 May 2022 Issuance of Amendment Nos. 314 and 204 Revise Technical Specifications to Adopt TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ML22082A2532022-03-29029 March 2022 Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27 ML22012A2972022-01-21021 January 2022 Proposed Alternative to Use ASME OM Code Case OMN-27 ML21197A0092021-11-0101 November 2021 Issuance of Amendment Nos. 313 and 203 Reactor Coolant System, Pressure and Temperature Limits Report ML21214A2752021-10-15015 October 2021 Issuance of Amendment Nos. 312 and 202 Atmospheric Dump Valves ML21153A1762021-06-30030 June 2021 Issuance of Amendment No. 201 Revision of Technical Specifications Related to Steam Generator Tube Inspection, and Repair Methods ML21075A1132021-04-16016 April 2021 Issuance of Amendment Nos. 311, 200, and 302 to Incorporate the Applicable Standard Technical Specification 5.2.2, Unit Staff, Into the Technical Specifications ML21070A0002021-03-22022 March 2021 Issuance of Amendment 310, Revise Technical Specification 5.5.5.1, Unit 1 SG Program, to Defer Spring 2021 Refueling Outage Steam Generator Inspections to Fall 2022 Refueling Outage ML20346A0222021-03-10010 March 2021 Issuance of Amendment Nos. 309 and 199 to Change Technical Specifications to Implement New Surveillance Methods for the Heat Flux Hot Channel Factor ML20335A0522021-02-18018 February 2021 Issuance of Amendment Nos. 308 and 198 to Modify Certified Fuel Handler Related Technical Specifications for Permanently Defueled Condition ML20345A2362021-01-28028 January 2021 Issuance of Amendment Nos. 307 and 197 to Add Containment Sump Technical Specifications to Address GSI-191 Issues ML20335A0232020-12-28028 December 2020 Issuance of Amendment Nos. 306 and 196 to Remove License Conditions B and C Related to the Irradiated Fuel Management Plans ML20285A2662020-10-21021 October 2020 Correction to Safety Evaluation for Amendment Nos. 305 and 195 Issued September 23, 2020, Modify Primary and Secondary Coolant Activity Technical Specifications ML20279A4402020-10-0808 October 2020 Energy Harbor Fleet-Beaver Valley Power Station; Davis-Besse Nuclear Power Station, and Perry Nuclear Power Plant - Request to Use a Provision of a Later Edition of the ASME Engineers Boiler and Pressure Vessel Code, Section XI ML20213A7312020-09-23023 September 2020 Issuance of Amendment Nos. 305 and 195 to Modify Primary and Secondary Coolant Activity Technical Specifications ML20223A0142020-08-20020 August 2020 Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19)) ML20206K8652020-08-0707 August 2020 Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19)) ML20153A0142020-07-16016 July 2020 Relief from the Requirements of the American Society of Mechanical Engineers Code Regarding Request VRR4 (EPID L-2020-LLR-0052 (COVID-19)) ML20145A0002020-06-0303 June 2020 Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19)) ML20080J7892020-04-28028 April 2020 Relief Requests 2 TYP-3-B3.110-1, 2-TYP-3-C2.21-1, 2-TYP-3-C1.30-1, and 2-TYP-3-RA-1 Regarding Weld Examination Coverage for the Third Inservice Inspection Interval ML20079F8162020-03-26026 March 2020 Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval ML19336A0282019-12-18018 December 2019 Safety Evaluation Irradiated Fuel Management Plans ML19305B1312019-12-0202 December 2019 Firstenergy Nuclear Operating Company - Enclosure 6, Non-Proprietary Safety Evaluation, Direct and Indirect Transfer of Licenses and Draft Conforming Amendments for Beaver Valley Units 1 and 2, Davis-Besse Unit 1, and Perry Unit 1 ML19275E2942019-10-16016 October 2019 Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements ML19028A0302019-04-11011 April 2019 FENOC - Beaver Valley Power Station, Unit 1 & 2; Davis-Besse Nuclear Power Station Unit 1; and Perry Nuclear Power Plant Unit 1 - Approval of Certified Fuel Handler Training and Retraining Program ML18348B2062019-02-25025 February 2019 Issuance of Amendment No. 193 Revise Steam Generator Technical Specifications ML18249A1692018-09-0707 September 2018 Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7800 and MF7801; EPID L-2016-JLD-0006) ML18227A7332018-08-27027 August 2018 Request for Relief from the Requirements of the ASME Code ML18205A2922018-08-10010 August 2018 Correction of Errors in Safety Evaluation Associated with Relief Request 2-TYP-4-RV-02 ML18179A4672018-07-30030 July 2018 FENOC-Beaver Valley Power Station, Unit No. 1 and 2; Davis-Besse Nuclear Power Station, Unit 1, and Perry Nuclear Power Plant, Unit 1 - Issuance of Amendments Request to Adopt TSTF-529, Clarify Use and Application Rules ML18178A1122018-07-0202 July 2018 Relief Request No. 2-TYP-4-RV-02 Regarding the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-4 Examination Requirements ML18065A4032018-04-0505 April 2018 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 301 and 190 ML18073A1062018-03-28028 March 2018 Safety Evaluation of Proposed Alternatives 1-TYP-4-BA-01 and TYP-4-BN-01 Regarding the Fourth 10- Year Interval of the Inservice Inspection Program ML18075A0962018-03-27027 March 2018 Safety Evaluation of Proposed Alternative to Use ASME Code Cases N-695-1 and N-696-1 in Lieu of Certain Requirements of the ASME Code 2023-09-28
[Table view] |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 28, 2015 Mr. Eric A. Larson, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NO.1- RELIEF REQUEST NO. 1TYP-4-RV-04 REGARDING THE EXAMINATION REQUIREMENTS OF CODE CASE N-729-1 (TAC NO. MF5049)
Dear Mr. Larson:
By letter dated October 17, 2014, 1 FirstEnergy Nuclear Operating Company (the licensee) submitted request 1TYP-4-RV-04 to the Nuclear Regulatory Commission (NRC). The licensee requested to use alternative requirements to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection (lSI) of Nuclear Power Plant Components," for Beaver Valley Power Station, Unit 1 (BVPS-1).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR)
Section 50.55a(a)(3)(i), (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested an alternative to performing the required volumetric and surface examinations for certain reactor pressure head components at the frequency prescribed in ASME Code Case N-729-1.
The NRC staff has reviewed the licensee's relief request and has determined that the requested alternative will provide an acceptable level of quality and safety, as documented in the enclosed safety evaluation. Therefore, the licensee's request for the use of the above stated alternative is authorized pursuant to 10 CFR 50.55a(z)(1) for the BVPS-1.
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 lnservice Inspector.
1 Agencywide Documents Access and Management System Accession No. ML14290A140.
E. Larson If you have any questions, please contact the Beaver Valley Project Manager, Taylor A. Lamb, at (301) 415-7128 or Taylor.Lamb@ nrc.gov.
Sincerely, Meena Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-334
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST REGARDING THE PROPOSED ALTERNATIVE TO ASME CODE CASE N-729-1 EXAMINATION FREQUENCY REQUIREMENTS FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION. UNIT 1 DOCKET NUMBER 50-334
1.0 INTRODUCTION
By letter dated October 17, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML14290A140), FirstEnergy Nuclear Operating Company (the licensee or FENOC) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, associated with the examination frequency requirements of Code Case N-729-1 for Beaver Valley Power Station, Unit No. 1 (BVPS-1 ).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR)
Section 50.55a(a)(3)(i) (retitled paragraph 50.55a(z)(1) by 79 FR 65776, dated November 5, 2014), the licensee requested to use the proposed alternatives in Relief Request 1TYP-4-RV-04, to the examination frequency of ASME Code Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1,"on the basis that the alternative examination provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(g)(6)(ii) state that "the Commission may require the licensee to follow an augmented inservice inspection [lSI] program for systems and components for which the Commission deems that added assurance of structural reliability is necessary."
The regulations in 10 CFR 50.55a(g)(6)(ii)(D), state, in part, that "all licensees of pressurized water reactors shall augment their [lSI] program with ASME Code Case N 729-1, subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section ... "
In this request, the licensee has requested relief from the examination frequency required by Code Case N-729-1 and has, therefore, also requested relief from 10 CFR 50.55a(g)(6)(ii)(D).
Enclosure
The regulations in 10 CFR 50.55a(z) state that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that (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.
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 Commission to authorize the proposed alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 COMPONENTS AFFECTED The affected components are ASME Class 1, Reactor Vessel Closure Head (RVCH)
Penetration Nozzle numbers 1 through 62, which are fabricated from lnconel SB-167 (Alloy 690)
UNS N06690. The nozzle J-groove welds are fabricated from UNS N06052 and UNS W86152, 52/152 weld materials. The original BVPS-1 RVCH penetration nozzles, which were manufactured with Alloy 600/82/182 materials, was replaced with a new RVCH using Alloy 690/52/152 material for the penetration nozzles during the refueling outage that returned to operation in April 2006.
3.2 INSERVICE INSPECTION INTERVAL The licensee's current lSI is the fourth 10-year lSI, which started on April 1, 2008, and ends on March 31, 2018. The proposed duration of the alternative is requested for the duration up to and including the 251h BVPS-1 refueling outage that is scheduled to commence in April2018.
3.3 ASME CODE OF RECORD The ASME Section XI Code of Record for the current, fourth 10-year lSI interval at BVPS-1, is the 2001 Edition through the 2003 Addenda.
3.4 ASME CODE AND/OR REGULATORY REQUIREMENTS Section 50.55a(g)(6)(ii)(D) of 10 CFR requires, in part, that licensees augment their lSI program in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D). ASME Code Case N-729-1, Table 1, Inspection Item 84.40 requires volumetric/surface examination be performed within one inspection interval (nominally 10 calendar years) of its in service date for a replaced RVCH. The required volumetric/surface examinations would thus have to be completed by early 2016 in order to fulfill the requirements of N-729-1.
3.5 PROPOSED ALTERNATIVE The licensee proposes to delay the next required inspection for a period of approximately 2 years. The licensee proposes to accomplish the inspection in accordance with ASME Code
Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D) during refueling outage 25, which is scheduled for April2018.
3.6 LICENSEE'S BASIS FOR USE OF THE PROPOSED ALTERNATIVE The licensee's basis for use of the proposed alternative is based primarily on three topics of consideration. The first topic addresses the concept that the inspection interval in Code Case N-729-1 is based on primary water stress-corrosion cracking (PWSCC) crack growth rates for Alloy 600/82/182. The second topic addresses a bare metal visual examination that was conducted on the licensee's replacement RVCH in 2010. The third topic addresses a plant-specific factor of improvement analysis that was conducted by the licensee.
In addressing its first basis for use of the proposed alternative, the licensee asserts that the inspection intervals contained in ASME Code Case N-729-1 for alloy 600/82/182 are based on re-inspection years ~RIY) equal to 2.25. This RIY value is based on PWSCC crack growth rates as defined in the 751 percentile curve contained in Materials Reliability Program (MRP) 55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," and MRP 115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds," both NRC approved documents. The licensee further asserts that the PWSCC crack growth rates of Alloy 690/52/152 are significantly lower than those of Alloy 600/82/182 and, therefore, merit a longer inspection interval. The licensee bases that assertion on: (1) the lack of cracking in other Alloy 690 components, such as steam generators and pressurizers, in the nearly 20 years that Alloy 690 has been in service in these components; (ii) the failure to observe cracking in inspections already performed in replacement heads (13 of 40 replacement heads in the United States have been examined, which includes heads that operate at higher temperatures than the head under consideration); (iii) the similarity of the inspected heads to the head under consideration regarding configuration, manufacturers, design and operating conditions; and (iv) laboratory test data for Alloy 690/52/152 as contained in MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles."
In addressing its second basis for use of the proposed alternative, the licensee stated that a bare metal visual examination was performed in 2010 on the BVPS-1 replacement RVCH in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This visual examination was performed by VT -2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any indications of nozzle leakage (e.g., boric acid deposits) on the surface or near a nozzle penetration. The licensee also indicated that this examination will be performed again in the upcoming refueling outage, which is scheduled to commence in April 2015. Also, the licensee stated that no alternative examination processes are proposed to those required by ASME Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The visual (VT-2) examinations and acceptance criteria, as required by Item B4.30 of Table 1 of ASME Code Case N-729-1, are not affected by this request and will continue to be performed on a frequency not to exceed every 5 calendar years.
In addressing its third basis for use of the proposed alternative, the licensee made a plant-specific calculation of the required factor of improvement in the crack growth rate of Alloy
690/52/152, as compared to the crack growth rate of Alloy 600/82/182. In making this calculation, the licensee used the actual temperature of the head and conservatively assumed that calendar years were equal to effective full-power years. Based on this calculation, the licensee determined that an improvement factor of 5.5 was required to meet the proposed and desired inspection interval of 12 calendar years. The licensee then proposed that, because the required factor of improvement (5.5) was smaller than the factor of improvement of 20, which bounded most of the MRP 375 data for Alloy 690/52/152, the use of a factor of improvement of 5.5 would not result in a reduction in safety and was, therefore, justified.
The licensee stated that their analysis showed significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the BVPS-1 replacement RVCH provide for a reactor coolant system pressure boundary, where the potential for PWSCC has been shown by analysis and by years of positive industry experience, to be remote. As such, the licensee found the technical basis sufficient to ensure public health and safety by extending the inspection frequency of the RVCH nozzle at BVPS-1 from a maximum of 10 years to a new maximum of 12 years.
3.7 NRC STAFF EVALUATION In evaluating the technical sufficiency of the licensee's proposed alternative (i.e., a one-time extension of the volumetric/surface examination interval contained in ASME Code Case N-729-1 from 10 years to no longer than 12 years), the NRC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative. The NRC staff found that the technical basis included by the licensee provided sufficient information for the NRC staff to review the proposed alternative.
Due to concerns about PWSCC, many PWR plants in the United States and overseas have replaced reactor vessel closure heads containing Alloy 600/182/82 nozzles with heads containing Alloy 690/152/52 nozzles. The inspection frequencies developed in Code Case N-729-1 for RVCH penetration nozzles using Alloy 600/182/82 were developed based, in part, on those material's crack growth rate equations, as documented in MRP 55 and MRP 115. The licensee's primary technical basis is to present crack growth rate data for the new, more crack resistant materials, Alloy 690/152/52, and demonstrate an improvement factor (IF) of these materials versus the older Alloy 600/82/182 materials. This IF would then provide the basis for the extension of the lSI frequency requested by the licensee in their proposed alternative.
In evaluating the licensee's first technical basis for use of the proposed alternative, the NRC staff notes that the licensee used MRP 375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles." This document, in part, summarizes numerous Alloy 690/152/52 crack growth rate data from various sources to develop IFs for the crack growth rate equations provided in MRP 55 and MRP 115. While the NRC staff finds the licensee's assertions and/or interpretations to be reasonable, MRP 375 is not an NRC approved document. As the NRC staff does not have sufficient time or resources to validate all of the data used by this document, the NRC staff does not consider it appropriate to use all of the data from this document to review the licensee's relief request. A more detailed review of the data provided in MRP 375 will be performed by an international group of experts as part of an Alloy 690 Expert Panel, which is currently scheduled to complete its review in the
2016-2017 timeframe. In the interim, the NRC staff review will rely upon Alloy 690/152/52 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data is documented in a data summary report and car:~ be found under ADAMS Accession Number ML14322A587. The NRC confirmatory research generally supports the contention that the crack growth rate of Alloy 690/52/152 is more crack resistant but differs from the MRP 375 data in some respects.
The PNNL and ANL data summary report includes crack growth rate data up to approximately 20 percent cold work based on the observation of local strains in welds and weld dilution zone data. However, the NRC staff did not consider the weld dilution zone data in its assessment.
This is because the limited weld dilution zone data that is currently available has shown higher crack growth rates than are commonly observed for Alloy 690/152/52 material. The high crack growth rates in weld dilution zones may be due to the reduced chromium present in these areas. The NRC staff chose to exclude the weld dilution zone data from this analysis due to the limited number of data points available, the variability in results, and due to the limited area of continuous weld dilution for flaws to grow through. For example, in the case of the highest measured crack growth rates, a flaw would have to travel in the heat affected zone of a j-groove weld along the low Alloy steel head interface. It is not fully apparent to the NRC staff how accelerated crack growth in very small areas of the weld dilution zone would result in a significantly increased probability of leakage or component failure during a relatively short extension of the required inspection interval. Exclusion of these data may be reevaluated as additional data become available; a better understanding of the existing data is obtained; or if a longer extension of the inspection interval is requested. Therefore, the NRC staff finds that the impact of these weld dilution zone crack growth rates on the change in volumetric inspection frequency, as requested by the licensee's proposed alternative, is not considered to be relevant for this specific relief request.
In evaluating the licensee's second basis for use of the proposed alternative, the NRC staff finds that the past bare metal visual examination on the head under consideration is a reasonable means to demonstrate the absence of leakage through the nozzle/J-groove weld, prior to the time the examination was conducted. The NRC staff also finds that performance of future bare metal visual examinations, in accordance with the code case, is adequate to demonstrate the absence of leakage at or prior to the time the examinations are conducted. Finally, the NRC staff finds that the proposed alternative's frequency for bare metal visual examinations, in conjunction with the new frequency of volumetric examinations, is sufficient to provide reasonable assurance of the structural integrity of the RVCH.
In evaluating the licensee's third basis for use of the proposed alternative, the NRC staff found that the licensee's calculated improvement factor of 5.5, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 12 calendar years, was found to be acceptable by NRC staff calculation. The NRC staff also found that the application of an IF of 5.5 to the 75th percentile curves in MRP 55 and 115 bounded essentially all of the NRC data included in the PNNL and ANL data summary report. Therefore, the NRC staff found that this analysis supports the concept that a volumetric inspection interval for the RVCH of not more than 12 calendar years does not pose a higher risk than that associated with an Alloy 600/182/82 RVCH inspected at intervals of 2.25 RIY. Hence, the NRC staff found the licensee's technical basis to be acceptable.
Therefore, based on the above evaluation, the NRC finds that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1)
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the alternative method proposed by the licensee in 1TYP-4-RV-04, Revision 0, will provide an acceptable level of quality and safety for the examination frequency requirements of the RVCH. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the one time use of 1TYP-4-RV-04 at Beaver Valley Power Station, Unit No.1, for the duration up to and including the 25th refueling outage that is scheduled to commence in April 2018.
All other requirements of the ASME Code,Section XI, and 10 CFR 50.55a(g)(6)(ii)(D) 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 Contributor: Steven Vitto Date: January 28, 2015
ML14363A409 *via e-mail OFFICE LPL 1-2/PM LPL 1-2/PM LPL 1-1/LA* DE/EMCB/BC* LPL 1-1/BC LPL 1-2/PM NAME TLamb JWhited A Baxter DAIIey MKhanna TLamb DATE 1/20/15 1/20/15 1/28/15 12/19/14 1/21/15 1/28/15