ML15349A956

Relief from the Requirements of the ASME Code

ML15349A956
Person / Time
Site:	Salem
Issue date:	12/24/2015
From:	Doug Broaddus Plant Licensing Branch 1
To:	Braun R Public Service Enterprise Group
Parker C, NRR/DORL/LPLI-2
References
CAC MF6089
Download: ML15349A956 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 24, 2015 Mr. Robert Braun President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (CAC NO. MF6089)

Dear Mr. Braun:

By letter dated April 8, 2015, PSEG Nuclear LLC (PSEG or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," requirements at Salem Nuclear Generating Station (Salem), Unit No. 1.

Specifically, PSEG proposes to extend the frequency of the volumetric/surface examination interval for the Salem, Unit No. 1, reactor vessel closure head (RVCH) nozzles and partial-penetration welds for approximately 5 years. Pursuant to Title 1O of the Code of Federal Regulations (10 CFR) Section 50.55a(z)(1 ), 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 has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the alternative method proposed by PSEG in Relief Request S1-14R-150 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 Relief Request S1-14R-150 at Salem, Unit No. 1, for the duration up to, and including, the 27th refueling outage that is scheduled to commence in fall 2020.

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.

R. Braun If you have any questions, please contact the Project Manager, Carleen Parker, at 301-415-1603 or Carleen.Parker@nrc.gov.

Sincerely,

-c.-~~

Douglas A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-272

Enclosure:

Safety Evaluation cc w/enclosure: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST S1-14R-150 REGARDING THE PROPOSED ALTERNATIVE TO ASME CODE CASE N-729-1 EXAMINATION FREQUENCY REQUIREMENTS PSEG NUCLEAR LLC SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NUMBER 50-272

1.0 INTRODUCTION

By letter dated April 8, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML15098A426), PSEG Nuclear LLC (PSEG or 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 for Salem Nuclear Generating Station (Salem),

Unit No. 1.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.55a(z)(1 ), the licensee requested to use the proposed alternatives in Relief Request S1-14R-150 to the examination frequency of 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, on the basis that the alternative examination provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

The inservice inspection (ISi) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," and applicable editions and addenda as required by 10 CFR 50.55a(g),

except where specific written relief has been granted by the U.S. Nuclear Regulatory Commission (NRC or the Commission).

Pursuant to 10 CFR 50.55a(g)(6)(ii), the Commission may require the licensee to follow an augmented ISi 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) require, in part, "[a]ll licensees of pressurized water reactors must augment their inservice inspection program with ASME Code Case N-729-1, subject to conditions specified in Enclosure

paragraphs (g)(6)(ii)(D)(2) through (6) .... " Relief from 10 CFR 50.55a(g)(6)(ii)(D), "Augmented ISi requirements: Reactor vessel head inspections," is requested by the licensee.

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 (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. The penetration nozzles are fabricated from lnconel SB-167 (Alloy 690) UNS N06690. The nozzle J-groove welds utilized ERNiCrFe-7 (UNS N06052) (Alloy 52) and/or ENiCrFe-7 (UNS W86152) (Alloy 152) weld materials.

The original Salem, Unit No. 1, RVCH that contained penetration nozzles manufactured with Alloy 600/82/182 materials was replaced with a new RVCH using Alloy 690/52/152 materials for the penetration nozzles and partial-penetration welds during the refueling outage preceding return to operation in November 2005.

3.2 ISi Interval The licensee's current ISi interval is the fourth 10-year ISi interval that started on May 20, 2011, and is scheduled to end on May 20, 2021. The proposed duration of the alternative is up to, and including, the Salem, Unit No. 1, refueling outage 1R27 that is scheduled to commence in fall 2020.

3.3 ASME Code of Record The ASME Section XI Code of Record for the current fourth 10-year ISi interval at Salem, Unit No. 1, is the 2004 Edition.

3.4 ASME Code and/or Regulatory Requirements Section 50.55a(g)(6)(ii)(D) of 10 CFR requires, in part, that licensees of pressurized water reactors (PWRs) shall augment their ISi program 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 1O calendar years) of its inservice date for a replaced RVCH. The required volumetric/surface examinations would need to be completed

during Salem, Unit No. 1, refueling outage 1R24, which is currently scheduled for spring 2016, in order to fulfill the requirements of N-729-1.

3.5 Proposed Alternative PSEG proposes to delay the next required volumetric/surface examination of the replacement RVCH for a period of approximately 5 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 27, which is scheduled for fall 2020.

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 conducted on the l_icensee's replacement RVCH in 2013. 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 value is based on PWSCC crack growth rates 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 Materials (MRP-55)," and in "Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115)" (MRP-55 and MRP-115 are available to the public at www.epri.com). 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 the following: (a) lack of cracking in other Alloy 690 components, such as steam generators and pressurizers, in the approximately 25 years that Alloy 690 has been in service in these components; (b) failure to observe cracking in inspections already performed in replacement heads (13 of 40 replacement heads in the United States have been examined, which include heads that operate at higher temperatures than the head under consideration),

including the Salem, Unit No. 2, RVCH, which was fabricated by the same manufacturer using the same nozzle material; (c) similarity of the inspected heads to the head under consideration regarding configuration, manufactures, design, and operating conditions; and (d) laboratory test data for Alloy 690/52/152 as contained in "Materials Reliability Program: 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 2010 and 2013 on the Salem, Unit No. 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 Salem, Unit No 1, refueling outage 1R25 that is scheduled to commence in fall 2017.

The visual (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, PSEG 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 6.21 was required to meet the proposed and desired inspection interval of 15 calendar years. The licensee then proposed that because the required factor of improvement (6.21) was smaller than the factor of improvement of 10, which bounds the MRP-375 data for Alloy 690/52/152 crack growth rates, the use of a factor of improvement of 6.21 would not result in a reduction in safety and was, therefore, justified.

PSEG states that its analysis shows significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the Salem, Unit No. 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 to be sufficient to provide an acceptable level of quality and safety for a one-time extension to the inspection frequency in accordance with 10 CFR 50.55a(z)(1 ).

3.7 NRG 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 1O years to not longer than 15 years), the N RC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative. The NRG staff found that the technical basis included by PSEG provided sufficient information for the NRG staff to review the proposed alternative.

Due to concerns about PWSCC, many PWR plants in the United States and overseas have replaced RVCHs 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 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 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 NRG staff notes that the licensee uses MRP-375. 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 NRG 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 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 that 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 can 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 materials. 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 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 this 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 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 6.21, to support an extension of the ASME

Code Case N-729-1 inspection frequency of 2.25 RIV to 15 calendar years, was acceptable by NRC staff calculation. The NRC staff also found that the application of an IF of 6.21 to the 751h 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 no more than 15 calendar years does not pose a higher risk than that associated with an Alloy 600/182/82 RVCH inspected at intervals of 2.25 RIV. Hence, the NRC staff found the licensee's technical basis to be acceptable.

Therefore, based on the above evaluation, the NRC staff finds that the proposed alternative provides an acceptable level of quality and safety as required by 10 CFR 50.55a(z)(1 ).

4.0 CONCLUSION

As set forth above, the NRC staff has determined that the alternative method proposed by PSEG in Relief Request S1-14R-150 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 Relief Request S1-14R-150 at Salem, Unit No. 1, for the duration up to, and including, the 271h refueling outage that is scheduled to commence in fall 2020.

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.

Principal Contributor: Robert H. Davis Date: December 24, 2015

R. Braun If you have any questions, please contact the Project Manager, Carleen Parker, at 301-415-1603 or Carleen.Parker@nrc.gov.

Sincerely, IRA REnnis for!

Douglas A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-272

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