ML14330A207

Relief Request ANO1-ISI-024, Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1, Fourth 10-year Inservice Inspection Interval

ML14330A207
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/23/2014
From:	Eric Oesterle Plant Licensing Branch IV
To:	Entergy Operations
George A
References
TAC MF4022
Download: ML14330A207 (9)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2014 Vice President, Operations

• Arkansas Nuclear One Entergy Operations.. Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 1 -REQUEST FOR ALTERNATIVE AN01-ISI-024 FROM VOLUMETRIC/SURFACE EXAMINATION FREQUENCY REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE CASE N-729-1 (TAC NO. MF4022)

Dear Sir or Madam:

By letter dated April28, 2014, as supplemented by letter dated October 2, 2014, Entergy Operations, Inc. (Entergy, the licensee), submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI requirements at Arkansas Nuclear One, Unit 1 (AN0-1).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i),

the licensee requested to use the proposed alternative to the examination frequency requirements of ASME Code Case N-729-1 for the reactor vessel closure head at AN0-1, on the basis that the proposed alternative provides an acceptable level of quality and safety.

The NRC staff has completed its review of the proposed alternative and based on the enclosed safety evaluation, the staff concludes that the alternative method proposed by the licensee in Relief Request AN01-ISI-024 will provide an acceptable level of quality and safety for the examination frequency requirements of the reactor vessel closure head. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements as set forth in 10 CFR 50.55a(a)(3)(i) for the proposed alternative. Therefore, the NRC staff authorizes the one-time use of AN01-ISI-024 at AN0-1 for the duration up to, and including the AN0-1 27th refueling outage that is scheduled to commence in April2018 and will occur in.the fifth 10-year inservice 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 and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.

If you have any questions, please contact the ANO Project Manager, Andrea George, at (301) 415-1081, or by e-mail at Andrea.George@nrc.gov.

Sincerely, Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

' ; .~ .. '

UNITED STATES NUCLEAR REGULATORY COMMISSION

'WAS.HII)IGTON, D.C. ~0555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE AN01-ISI-024 FROM VOLUMETRIC/SURFACE EXAMINATION FREQUENCY REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE CASE N-729-1 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313

1.0 INTRODUCTION

• ..

• By letter dated April 28, 2014 (Agencywide Documents Access and Management System

• * (ADAMS) Accession No. ML14118A477), as supplemented by letter dated October 2, 2014 (ADAMS Accession No. ML14275A460), Entergy Operations, Inc. (Entergy, the licensee),

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, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1,"at Arkansas Nuclear One, Unit 1 (AN0-1).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i),

the licensee requested to use the proposed alternative in Relief Request AN01-ISI-024, 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 inservice inspection (lSI) 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),

"lnservice Inspection requirements," except where specific written relief has been granted by the Commission.

The regulations in 10 CFR 50.55a(g)(6)(ii) state that "[t]he Commission may require the licensee to follow an augmented inservice inspection 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, that "[a]lllicensees of pressurized water reactors Enclosure

shall augment their i!lservice inspection program With ASME Code Case N-729-1, subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) .... "

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the requirements of 10 CFR 50.55a(g), may be used when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates that (i) the proposed alternatives would provide an acceptable level of quality and*safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.0 TECHNICAL EVALUATION

3.1 Components Affected The licensee stated that the affected components are ASME Code Class 1, Reactor Vessel Closure Head (RVCH) Penetration Nozzles 0-1 through 0-69, 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), Alloy 52/152 weld materials. The licensee also stated that the original AN0-1 RVCH penetration nozzles, which were manufactured with .

Alfoy 600/82/182 materials, were replaced with a new RVCH using Alloy 690/52/152 material for the penetration nozzles, which was placed in service in December 2005.

3.2 lnservice Inspection Interval AN0-1 is currently in its fourth 10-year lSI interval (May 31, 2008, through May 30, 2017). The NRC staff notes that the proposed duration of the alternative would end in the fifth 10-year lSI interval, which is from May 31, 2017, to May 30, 2026.

3.3 ASME Code of Record The llicensee stated that the ASME Section XI Code of record for the current fourth 10-year lSI interval at AN0-1, which began on May 31, 2008, and ends on May 30, 2017, 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, licensees to augment their lSI programs 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 inservice date for a replaced RVCH. The required volumetric/surface examinations would thus have to be completed by December 2015 in order to fulfill the requirements of ASME Code Case N-729-1.

3.5 Proposed Alternative The licensee proposed to delay the next required volumetric/surface examination of the replacement RVCH for a period of approximately 2.5 years from its current inspection date.

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 the AN0-1 27th refueling outage, which is scheduled for April 2018. The NRC staff notes that the current required inspection date occurs in the plant's fourth lSI interval, and that the proposed inspection will be accomplished during the plant's fifth lSI interval.

3.6 Licensee's Basis for Use of the Proposed Alternative The licensee's basis for the proposed alternative is comprised of the following: (1) the inspection interval in ASME Code Case N-729-1 is based on primary water stress-corrosion cracking (PWSCC) crack growth rates for Alloy 600/82/182', which are conservative compared to the lower crack growth rates for Alloy 690/52/152; (2) bare metal visual examination conducted on the licensee's replacement RVCH in 201 0; and (3) a plant-specific factor of improvement analysis conducted by the licensee.

In addressing its first basis for use of the proposed alternative, the licensee stated 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. This RIY value was developed based on PWSCC crack growth rates as defined in the 75th percentile curve contained in Electric Power Research Institute (EPRI) Materials Reliability Program (MRP)-55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," and EPRI MRP-115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds" (MRP-55 and MRP-115 are available to the public at www.epri.com). The licensee stated that the Alloy 690/52/152 replacement RVCH has inc-reased resistance to PWSCC over that of Alloy 600/82/182 and, therefore, a limited extension of volumetric/surface examination frequency is acceptable. The licensee bases this assertion on: 1) industry operating experience showing resistance to PWSCC for Alloy 690 components, such. as steam generators and pressurizers in the approximately 20 years that Alloy 690 has been in service in these components; 2) particularly, the lack of observed cracking in lSI volumetric/surface examinations of 9 of 40 RVCHs in the United States (some of which had continuous full..:power operating temperatures approaching 613 degrees Fahrenheit, which bounds the RCVH operating temperature for AN0-1 ); 3) the similarity of other RVCHs to the AN0-1 replacement RVCH under consideration, regarding configuration, design, and operating conditions (including Oconee Nuclear Station, Units 1, 2, and 3, and Three Mile Island Nuclear Station, Unit 1, all of which completed volumetric/surface.examinations, and did not reveal any recordable indications); and 4) laboratory test data for Alloy 690/52/152 as contained in EPRI MRP-375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles" (available to the public at www.epri.com).

In addressing its second basis for use of the proposed alternative, the licensee stated the following in its letter dated April 28, 2014:

A bare metal visual examination was performed in 2010 on the AN0-1 replacement RVCH in accordance with ASME Code Case N-729-~, 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 surface or nozzle penetration boric acid that would be indicative of nozzle leakage. This examination will be

performed again in the upcoming 25th refueling outage scheduled to commence in January 2015.

In its letter dated April 28, 2014, the licensee also stated, in part, that No alternative examination processes are proposed to those required by ASME Code Case N-729-1, as conditioned by 10CFR50.555a(g)(6)(ii)(D). The visual

{VT-2) examinations and acceptance criteria as required by Item 84.30 of Table 1 of ASMECode 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 described a plant-specific calculation of the required factor of improvement (FOI) in the crack growth rate of Alloy 690/52/152 as compared to the crack growth rate of Alloy 600/82/182. As inputs to the calculation, the licensee used the AN0-1 operating temperature of the RVCH and a conservative .assumption that effective full-power years, since RVCH is equal to the calendar years. Based on this calculation, the licensee determined a new RIY value of 17.24 and a new FOI of 7.7. The licensee stated that the new FOI implied by the requested extension period .

represents a level of reduction in PWSCC crack growth rate versus that for Alloy 600/82/182 that is bounded on a statistical basis by the data compiled in EPRI MRP-375.

In its letter dated April 28, 2014, the licensee stated, in part, that "the Alloy 690 nozzle base and Alloy 52/152 weld materials used in the AN0-1 replacement RVCH provide for a clearly superior 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." The licensee also stated that "the AN0-1 RVCH FOI corresponding to the requested period of extension to perform a volumetric/surface examination provides an acceptable level of quality and safety in accordance with 10CFR50.55a(a)(3)(i)."

3.7 NRC Staff Evaluation In evaluating the technical sufficiency of the licensee's proposed alternative, the NRC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative.

Due to concerns regarding PWSCC, many PWR plants in the United States and overseas have replaced RVCHs containing Alloy 600/182/82 nozzles with RVCHs containing Alloy 690/152/52 nozzles. The inspection frequencies developed in ASME Code Case N-729-1 for RVCH '

penetration nozzles using Alloy 600/182/82 were based, in part, on those material's crack growth rate equations documented in EPRI MRP-55 and EPRI MRP-115. The licensee's application, as supplemented, provided information and data regarding the more PWSCC-resistant materials, Alloy 690/152/52, and calculations to demonstrate an improvement factor (IF) of these materials versus the Alloy 600/82/182 materials. This IF would then provide the basis for the e?<tension of the lSI 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 EPRI MRP-375. This document, in part, summarizes Alloy 690/52/152 crack growth rate data from various sources to develop IFs for the crack

• growth rate equations provided in EPRI MRP-55 and EPRI MRP-115. While the NRC staff finds

that the licensee's assertions *and/or interpretations regarding EPRI MRP-375 are reasonable, EPRI MRP-375 is not an NRC-approved document. Therefore, since the licensee did not request review and approval of EPRI MRP-375 for this proposed alternative, the NRC staff did not use the data from this document to review the licensee's relief request. A detailed review of the data provided in EPRI 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 wil.l rely upon Alloy 690/152/52 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). The data from these two contractors is documented in a data summary report and can be found under ADAMS Accession No. ML14322A587. This confirmatory research regarding Alloy 690/52/152 crack growth rates, performed by PNNL and ANL, generally supports the information provided by the licensee in its application, as supplemented, regarding the assertion that Alloy 690/52/152 are more crack-resistant than Alloy 600/82/182, but differs from the EPRI MRP-375 crack growth rate 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 review pertaining to* the licensee's proposed alternative. 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 690/52/152 materials. The high crack growth rates in weld dilution zones may be due to the reduced chromium present in these areas. The NRC staff excluded the weld dilution zone data from this analysis due to the limited number of data points available, the variability in results, and the limited area of continuous weld dilution through which flaws grow.

For example, in the case of the highest measured crack growth rates, a flaw would have to travel iri 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 becomes 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 concludes 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 concludes 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 concludes 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 concludes 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 concludes that the licensee's calculated IF of 7. 7, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 12.5 calendar years, is acceptable by NRC staff calculation. The NRC staff also concludes that the application of an IF of 7.7 to the 75 1h percentile curves in EPRI MRP-55 and EPRI MRP-115 bounded essentially all of the NRC data included in the PNNL and ANL data summary report. Therefore, the NRC staff concludes that this analysis supports the licensee's assertion that a volumetric inspection interval for the AN0-1 replacement RVCH of not more than 12.5 calendar years and does not pose a higher risk than that associated with an Alloy 600/182/82 RVCH inspected at intervals of 2.25 RIY.

Therefore, the NRC staff concludes that the licensee's technical basis for the proposed alternative is acceptable.

4.0 CONCLUSION

As set forth above, the NRC staff concludes that the proposed alternative method by the licensee in Request for Relief AN01-ISI-024 will provide an acceptable level of quality and safety for the examination frequency requirements of the reactor vessel closure head.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth fn 10 CFR 50.55a(a)(3)(i) for the proposed alternative.

Therefore, the NRC staff authorizes the one-time use of AN01-ISI-024 at AN0-1, for the duration up to and including the AN0-1 27th refueling outage, which is scheduled to commence in April 2018, and will occur in the fifth 10-year lSI inspection interval.

All other requirements of theASME 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: S. Vitto, NRR/DE/EPNB Date: December 23, 2014

ML14330A207 *concurrence via email OFFICE NRR/DORLILPL4-1 /PM NRR/DORLILPL4-2/LA NRR/DORL/LPL4-1 /LA NRR/DE/EPNB/BC* NRR/DORLILPL4-1 /BC(A)

NAME AGeorge PBiechman JBurkhardt DAiley (JTsao for) EOesterle DATE 12/15/14 12/10/14 12/15/14 11/17/14 12/23/14