ML15021A354
| ML15021A354 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/17/2015 |
| From: | Shana Helton Plant Licensing Branch II |
| To: | Glover R Duke Energy Carolinas |
| Barillas M DORL/LPL2-2 301-415-2760 | |
| References | |
| TAC MF4801 | |
| Download: ML15021A354 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Richard Michael Glover Site Vice President H. B. Robinson Steam Electric Plant Duke Energy 3581 West Entrance Road Hartsville, SC 29550 February 17, 2015
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - RELIEF REQUEST NO. 11 - RELIEF FROM VOLUMETRIC/SURFACE EXAMINATION FREQUENCY REQUIREMENTS OF ASME CODE CASE N-729-1 (TAC NO. MF4801)
Dear Mr. Glover:
By letter dated August 27, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14251A014), as supplemented by letter dated November 10, 2014 (ADAMS Accession No. ML14325A693), Duke Energy Progress, Inc., (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," associated with the examination frequency requirements of Code Case N-729-1 at H. B. Robinson, Unit No. 2 (RNP-2).
Specifically, the licensee proposed to extend the frequency of the volumetric/surface examinations of the reactor vessel closure head (RVCH) nozzles and partial-penetration welds for approximately 3 years beyond the one inspection interval (nominally 10 calendar years) from the installation of the RNP replacement RVCH. Pursuant to Title 10 of the Code of Federal Regulations (10 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 alternative on the basis that the alternative provides an acceptable level of quality and safety.
The NRC staff reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). Therefore, the NRC staff authorizes the one-time use of Relief Request No. 11 at RNP for the duration up to, and including, the 31 51 RNP refueling outage that is scheduled to commence in September of 2018, and which will occur in the fifth 10-year inservice inspection interval, which began July 22, 2012, and ends July 30, 2021.
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.
If you have any questions or concerns, please contact the Project Manager, Martha Barillas, at 301-415-2760, or by e-mail at Martha.Barillas@nrc.gov.
Docket No. 50-261
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via Listserv Sincerely, Shana R. Helton, Branch Chief Plant Licensing Branch 11-2 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 RELIEF REQUEST NO. 11 REGARDING INSPECTION OF REACTOR VESSEL CLOSURE HEAD NOZZLES DUKE ENERGY PROGRESS, INC.
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
By letter to the U.S. Nuclear Regulatory Commission (NRC or Commission) dated August 27, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14251A014), as supplemented by letter dated November 10, 2014 (ADAMS Accession No. ML14325A693), Duke Energy (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, associated with the examination frequency requirements of Code Case N-729-1 at H. B. Robinson, Unit 2 (RNP).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 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 alternative to the examination frequency of ASME Code Case N-729-1, on the basis that the alternative examination provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The regulation in 1 O CFR 50.55a(g)(6)(ii), states, in part, that the Commission may require the licensee to follow an augmented inservice inspection (ISi) program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.
Further, 10 CFR 50.55a(g)(6)(ii)(D), states, in part, that all licensees of pressurized water reactors (PWRs) shall augment their ISi program with ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section.
Enclosure In Relief Request (RR) No. 11, the licensee requests 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).
The regulations of 10 CFR 50.55a(z)(1) and 50.55a(z)(2) 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 nozzles and partial penetration welds, which are fabricated from lnconel SB-167 (Alloy 690)
UNS N06690. The nozzle J-groove welds are fabricated from ERNiCrFe-7 (UNS N06052) and ENiCrFe-7 (UNS W86152) 52/152 weld materials. The original RNP RVCH that contained penetration nozzles, which was 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 October 2005.
3.2 lnservice Inspection Interval The proposed duration of the alternative will occur in the fifth 10-year ISi interval, which began July 22, 2012, and ends July 30, 2021.
3.3 ASME Code of Record The ASME Code of Record for the fifth 10-year ISi interval is the 2007 Edition with 2008 Addenda of ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components."
3.4 ASME Code and/or Regulatory Requirements Section 50.55a(g)(6)(ii)(D) of 1 O CFR requires, in part, that licensees shall augment their ISi program in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 1 O 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 inservice date for a replaced RVCH. The required volumetric/surface examinations would thus have to be completed by October 2015 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 3 years beyond the one inspection interval from installation of the RNP replacement RVCH.
The licensee proposes to accomplish the inspection in accordance with ASME Code Case N-729-1and10 CFR 50.55a(g)(6)(ii)(D) during Refueling Outage 31, which is scheduled for September 2018.
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 (CGRs) for Alloy 600/82/182. The second topic addresses a bare metal visual examination conducted on the licensee's replacement RVCH in 2010. The third topic addresses a plant-specific factor of improvement (FOi) analysis 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 reinspection years (RIY) equal to 2.25 and that this value is based on PWSCC CGRs as defined in the 751h percentile curve contained in "Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material (MRP-55)," and "Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115NP)." The licensee further asserts that the PWSCC CGRs 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 (a) the lack of cracking in other 690 components, such as steam generators and pressurizers, in the approximately 20 years that Alloy 690 has been in service in these components; (b) the failure to observe cracking in inspections already performed in replacement heads (9 of 40 replacement heads have been examined, which includes heads that operate at higher temperatures than the head under consideration); (c) the similarity of the inspected heads to the head under consideration regarding configuration, manufacture, design, and operating conditions; and ( d) laboratory test data for Alloy 690/52/152, as contained in "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375)."
In addressing its second basis for use of the proposed alternative, the licensee stated that a bare metal visual examination was performed in 201 O on the RNP replacement RVCH in accordance with ASME Code Case N-729-1, Table 1, Item 84.30. 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 29th refueling outage scheduled to commence in May 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 VT-2 examinations and acceptance criteria as required by Item 84.30 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 FOi in the CGR of Alloy 690/52/152, as compared to the CGR 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 (IF) of 5.7 was required to meet the proposed and desired inspection interval of 13 calendar years. The licensee then proposed that because the required FOi (5. 7) was smaller than the FOi of 20, which bounded most of the MRP 375 data for Alloy 690/52/152, the use of an FOi of 5. 7 would not result in a reduction in safety and was, therefore, justified.
The licensee stated that its analysis showed significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the RNP 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 RNP from a maximum of 10 years, to a new maximum of 13 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 13 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 Stated 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 materials' CGR equations documented in MRP-55 and MRP-115. The licensee's primary technical basis is to present CGR data for the new more crack-resistant materials, Alloy 690/152/52, and demonstrate an IF of these materials versus the older Alloy 600/82/182 materials. This IF would then provide the basis for the extension of the ISi frequency requested by the licensee in its proposed alternative.
In evaluating the licensee's first technical basis for use of the proposed alternative, the NRC staff notes that the licensee uses MRP-375. This document, in part, summarizes numerous Alloy 690/152/52 CGR data from various sources to develop IFs for the CGR 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 NRG-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 RR. 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 uses Alloy 690/152/52 CGR data from two NRC contractors:
Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data is documented in a data summary report dated October 30, 2014, and can be found under ADAMS Accession No. ML14322A587. The NRC confirmatory research generally supports the contention that the CGR 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 CGR data up to approximately 20 percent cold work based on the observation of local strains in welds and weld dilution zone data (WDZD). However, the NRC staff did not consider the WDZD in its assessment. This is because the limited WDZD that is currently available has shown higher CGRs than are commonly observed for Alloy 690/152/52 material. The high CGRs in weld dilution zones may be due to the reduced chromium present in these areas.
The NRC staff chose to exclude the WDZD from this analysis due to the limited number of data points available, the variability in results, and the limited area of continuous weld dilution for flaws to grow through. For example, in the case of the highest measured CG Rs, 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 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 CG Rs on the change in volumetric inspection frequency, as requested by the licensee's proposed alternative, is not considered to be relevant for this specific RR.
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 IF of 5.7, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 13 calendar years, was acceptable by NRC staff calculation.
The NRC staff found that the application of an IF of 5.7 to the 75th percentile curves in MRP-55 and MRP-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 13 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.
The NRC staff found the licensee's technical basis to be acceptable.
Therefore, the NRC finds that the proposed alternative provides an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(z)(i).
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the alternative method proposed by the licensee in RR No. 11 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(a)(z)(i) and is in compliance with ASME Code requirements. Therefore, the NRC staff authorizes the one-time use of RR No. 11 at RNP, Unit 2, for the duration up to, and including, the 31st refueling outage that is scheduled to commence in September 2018 and which will occur in the fifth 10-year ISi inspection interval.
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 approved remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: Margaret T. Audrain Date: February 17, 2015
ML15021A354 RidsNrrDeEpnb LRonewicz, NRR RidsNrrDorlDpr
- via memo OFFICE NRR/DORULPL2-2/PM NRR/DORULPL2-2/LA (LAiT)
NRR/DORULPL2-2/LA NAME MBarillas LRonewicz BClayton DATE 02/02/15 02/03/15 02/04/15 OFFICE NRR/DE/EPNB*
NRR/DORULPL2-2/BC NRR/DORULPL2-2/PM NAME DAiiey SHelton MBarillas DATE 12/18/14 02/11/15 02/17/15