ML16013A348

Alternative to Inservice Inspection

ML16013A348
Person / Time
Site:	Farley
Issue date:	02/01/2016
From:	Markley M Plant Licensing Branch II
To:	Pierce C Southern Nuclear Operating Co
Williams S
References
CAC MF6475
Download: ML16013A348 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 1, 2016 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

P. 0. Box 1295 /Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

JOSEPH M. FARLEY, UNIT 1, ALTERNATIVE TO INSERVICE INSPECTION (CAC NO. MF6475)

Dear Mr. Pierce,

By letter dated July 17, 2015, as supplemented by letter dated November 25, 2015, Southern Nuclear Operating Company, Inc., (SNC) requested a one-time extension of the inservice inspection (ISi) interval for examinations of the reactor pressure vessel (RPV) welds (Category B-A) and the nozzle-to-vessel welds and inner radius sections (Category B-D) from 1O to 20 years.

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 alternative on the basis that the alternative provides an acceptable level of quality and safety.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request, and concludes that SNC has adequately addressed all of the regulatory requirements and that the proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes the proposed alternative in accordance with 10 CFR 50.55a(z)(1 ), The NRC staff's:safety ~valuation is enclosed.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third party review by the Authorized Nuclear.lnservice Inspector.

C. Pierce If you have any questions, please contact the Project Manager, Shawn Williams, at 301-415-1009 or via e-mail at Shawn.Williams@nrc.gov.

Sincerely, Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-348

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

L!NITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 SOUTHERN NUCLEAR OPERATING COMPANY ALTERNATIVE FNP-ISl-ALT-18, VERSION 1.0 DOCKET NO. 50-348

1.0 INTRODUCTION

By letter dated July 17, 2015, as supplemented by letter dated November 25, 2015, (Agencywide Document Access and Management System (ADAMS) Accession Nos.

ML15198A155 and ML15329A212, respectively), Southern Nuclear Operating Company (SNC, the licensee) proposed an extension of the inservice inspection (ISi) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI, ISi Program for Farley Nuclear Plant, Unit 1 (FNP-Unit 1). Enclosure 1 of the submittal contains Alternative FNP-ISl-ALT-18, requesting to use an alternative to the requirements of the ASME Code,Section XI, Table IWB-2500-1.

Specifically, FNP-ISl-ALT-18, Version 1.0, proposes an alternative pursuant to Section 50.55a(z)(1) of Title 10 of the Code of Federal Regulations (10 CFR) to extend the ISi interval for examinations of the reactor pressure vessel (RPV) welds (Category B-A) and the nozzle-to-vessel welds and inner radius sections (Category B-D) from 10 to 20 years. FNP-Unit 1 is currently in the 4th 10-year ISi interval dated December 1, 2007, through November 30, 2017.

2.0 REGULATORY EVALUATION

2.1 Regulations and Guidance In accordance with 10 CFR 50.55a(g)(4), the licensee is required to perform ISi of ASME Code Class 1, 2, and 3 components and system pressure tests during the first 10-year interval and subsequent 10-year intervals that comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10*CFR 50.55a(a)(1 )(ii),

subject to the limitations and modifications listed in 10 CFR 50.55a(b)(2).

Enclosure

For the fourth ten-year ISi interval at FNP-Unit 1, the Code of record for the inspection of ASME Code Class 1, 2, and 3 components is the 2001 Edition through the 2003 Addenda of the ASME Code,Section XI. The regulation in 10 CFR 50.55a(z)(1) states, in part, that the Director of the Office of Nuclear Reactor Regulation may authorize an alternative to the requirements of 10 CFR 50.55a(g). There are two justifications for an alternative to be authorized. First, per 10 CFR 50.55a(z)(1 ), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. For the second possible justification for an alternative to be authorized, described in 10 CFR 50.55a(z)(2), the licensee must show that following the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials," describes general procedures acceptable to the staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled RPVs.

RG 1.174, Rev. 1, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes methods and assumptions acceptable to the staff for determining the RPV neutron fluence.

2.2 Background The ISi of Categories B-A and B-D components consists of periodic volumetric examinations intended to discover whether flaws have initiated, whether pre-existing flaws have extended, and whether pre-existing flaws may have been missed in prior examinations. These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code.

2.3 Summary of Technical Basis WCAP-16168-NP-A, Rev. 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (ML11306A084, referred to as the WCAP-A in the rest of this SE) was issued by the Pressurized Water Reactor Owners Group (PWROG). The report took data associated with three different PWR plants (referred to as the pilot plants), one designed by each of the three main vendors (Westinghouse, Combustion Engineering .(CE), and Babcock and Wilcox (B&W)) for domestic pressurized water reactor (PWR) nuclear power plants, and performed studies on these pilot plants to justify the proposed extension of the ISi interval for Categories B-A and B-D components from 10 to 20 years. The analyses incorporated the effects of fatigue crack growth and ISi. From the results, the PWROG concluded that the ASME Code,Section XI 10-year inspection interval for Categories B-A and B-D components in PWR RPVs can be extended to 20 years. Their conclusion from the results for the pilot plants was considered to apply to any plant designed by the three vendors as long as the critical, plant-

specific parameters (defined in Appendix A of WCAP-16168-NP-A, Rev. 2) are bounded by the pilot plants.

The staff's SE indicates that the methodology presented in WCAP-A is consistent with RG 1. 174, Rev. 1, and is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and .

conditions in the SE. In addition to showing that the subject plant parameters and inspection history are bounded by the critical parameters identified in Appendix A in WCAP-A, the licensee's application must provide the following plant-specific information:

(1) licensees must demonstrate that the 95th percentile total through-wall cracking frequency (TWCFTOTAL) of their RPV is within the envelope used in the supporting analyses. The 95th percentile TWCFTOrAL must be calculated using the methodology in NUREG-1874. The RT MAx-x and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, lff30, must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRG-approve~

methodology.

(2) Licensees must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in WCAP-A that are considered to significantly contribute to fatigue crack growth.

(3) Licensees must report the results of prior ISi of RPV welds and the proposed schedule for the next 20-year ISi interval. The 20-year inspection interval is a maximum interval.

In its request to implement the alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-10-238 (ML11153A033). .

(4) Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bounds the fatigue crack growth for all of its design basis transients and (b) identify the design basis transients that contribute to significant fatigue crack growth.

(5) RP.Vs with forgings that are susceptible to underclad cracking and with maximum Reference Temperature (RTMAx-Fo) values exceeding 240°F are not bounded by the

• WCAP-A. To extend the inspection interval for ASME Code,Section XI, Category B-A and B-D RPV welds from 1O to a maximum of 20 years, a licensee will have fo submit a plant-specific evaluation.

(6) Licensees seeking second or additional interval extensions shall provide the information and analyses requested in s.ection (e) of 10 CFR 50.61a. *

3.0 TECHNICAL EVALUATION

3.1 Proposed Alternative 3.1.1 Description of Proposed Alternative In FNP-ISl-ALT-18, the licensee proposed to defer the ASME Code required Categories B-A and 8-D volumetric examinations currently scheduled for the Fall of 2016 (3rd period of the 4th ISi interval) until no later than the end 9f November 2027 (3rd period of the 5th ISi interval).

3.1.2 Components for Which the Alternative is Requested The affected components are the subject plant RPV and their interior attachments and core support structures. The following examination categories and item numbers from IW8-2500 and Table IW8-2500-1 of the ASME Code,Section XI, are addressed in this alternative:

Exam Category Item Number Description 8-A 81.11 Circumferential Shell Welds 8-A 81.12 Longitudinal Shell Welds 8-A 81.21 Circumferential Head Welds 8-A 81.22 Meridional Shell Welds 8-A 81.30 Shell-to-Flange Weld 8-D 83.90 Nozzle-to-Vessel Welds 8-D 83.100 Nozzle Inside Radius Section 3.1.3 Basis for Proposed Alternative The licensee stated that the methodology used to demonstrate the acceptability of extending the inspection interval for Examination Category 8-A and 8-D components is contained in the WCAP-A. This methodology.used the estimated TWCF as a measure of the risk of RPV failure, and it was demonstrated that the inspection interval for the affected components can be extended from 10 to 20 years, meeting the change in risk guidelines in RG 1.174. The licensee addressed the plant-specific information discussed in Section 2.4 of this SE as follows:

{1) A plant-specific analysis, with identified critical parameters and detailed TWCF calculation demonstrated that the FNP-Unit 1 RPV's parameters are bounded by corresponding pilot plant parameters. The total TWCFs were calculated as 4.81 E-11 for FNP-Unit 1, less than the value of 1. 76E-08 for the Westinghouse pilot plant in the WCAP-A.

(2) The frequencies of the FNP-Unit 1 RPV's limiting design basis transients are bounded by the frequencies identified in the PWROG fatigue analysis.

(3) The results of the previous RPV inspection for the FNP-Unit 1 are provided. Three subsurface indications were found in plate material near the subject welds from the beltline region; all were within the inner 1/10th or 1 inch of the vessel wall thickness and were acceptable according to Table IW8-3510-1 of Section XI of the ASME Code. The largest indication had a through-wall extent of 0.38 inches, exceeding the requirements of the Alternate PTS Rule, 10 CFR 50.61 a. The one large indication is not considered

significant because the plate where the flaw is located is not the limiting beltline material, the total number of indications is far le~s than what is allowed in the Alternate* PTS Rule, and the magnitude of the calculated TWCF95-TOTAL is more than 2 orders of magnitude less than that for the pilot plant RPV in WCAP-A. The licensee also noted that a larger plate flaw was found in a different domestic unit, but the presence of the large flaw did not prevent approval of the same alternative for that unit. The RPV examination currently scheduled for 2016 is proposed to be deferred until the end of November of 2027. The date provided is consistent with the date identified in PWROG letter OG-10-238, dated July 12, 2010.

Plant-specific information items (4), (5), and (6) have not been addressed by the licensee because they do not apply to FNP-Unit 1. Since FNP-Unit 1 is bounded by the pilot plant application, the licensee concluded that use of this proposed alternative will provide an acceptable level of quality and safety and requested, pursuant to 10 CFR 50.55a(z)(1), that the NRC authorize the alternative.

• 3.1.4 Duration of Proposed Alternative The fourth ISi interval for the ASME Categories B-A and B-D RPV welds currently ends at the end of November 2017. The proposed alternative would require the inspections associated with the fourth ISi interval be done before the end of November 2027.

3.2 NRC Staff Evaluation The NRC staff reviewed the alternative where the licensee proposes to defer the ASME Code required volumetric examination of the FNP-Unit 1 RPV full penetration pressure-retaining Examination Category B-A and B-D welds for the fourth ISi interval. The inspections are currently scheduled for the fall of 2016, and instead, the alternative proposes to inspect the subject welds no later than the end of November in 2027.

To address plant-specific information item* 1, the licensee performed the TWCF calculations using the WCAP-A methodology with RG 1.99, Revision 2 to account for neutron embrittlement for each beltline region. The NRC staff notes that WCAP-A specifically states that any NRC-approved method to account for neutron embrittlement can be used with the WCAP-A analysis.

The NRC staff performed independent TWCF calculations that used the inputs from Table 3 of the alternative and two different methods to estimate .liT3o. The first method used RG 1.99, Rev. 2, Position 1.1 and Position 2.1 when credible surveillance data was available. The second method to calculate AT 30 used NUREG-187 4, as described in WCAP-A. The NRC staff's calculated 1WCF95~TOTAL values are nearly the same as that of the licensee's, and all of the plant-specific values are several orders of magnitude lower than the value for the Westinghouse pilot plant in the WCAP-A. Therefore, the NRC staff found the analysis to be acceptable.

For plant-specific information 2, the NRC staff confirmed the "Freque_ncy and Severity of Design

• Basis Transients" of FNP-Unit 1 is below the number in the conditions and limitations described in the WCAP-A in Table 1 of the alternative.

In the NRC staff's review of plant-specific information item 3 (information pertaining to previous RPV inspection and the schedule for future ones), the NRC staff considered the licensee's conclusion that the largest indication is acceptable even though it exceeded the requirements for plate material in 10 CFR 50.61a. In addition to the issues noted by the licensee, the NRC staff notes that that same indication would be acceptable per Table 3of10 CFR 50.61a if the indication were present in the weld rather than the plate. Furthermore, the NRC staff agrees with the licensee that the previous experience at Calvert Cliffs, Unit 2, is bounding for the one large plate indication, but with one exception; the submittal does not include any history of earlier inspections of that region of the RPV or the specific location of the indications with respect to the cladding base-metal interface.

To better characterize the indications found in the third ISi, the NRC staff requested additional information regarding the location and size of the indications that were found in the RPV beltline area, and specifically whether the indications were observed in any of the previous inspections. The RAI also asked whether the size of the indication changed during the course of the three inspections, and if so, whether the size of the indication could be attributed to improved inspection procedures.

By letter dated November 25, 2015, the licensee responded to the NRC staffs RAI. The response summarized the methods that were used during the previous inspections and discussed the findings for the two most recent set of exams. The most recent inspections performed for the third interval used a contact technique from the inside surface of the RPV, and the exams found the three indications described in the submittal. The exams *utilized a technique qualified under ASME Section XI Appendix VIII. The sizing techniques of ASME Section XI, Appendix VIII, provide the most reliable sizing techniques, as these examinations require specific examination standards, which included a rigorous qualification process.

The exams for the second interval were performed to ASME Section XI, which used Article 4 of the ASME Code,Section V, as the technique basis. This examination was performed from the inside*

surface of the RPV; seven indications were found and compared to a 2% (t) notch.

The decrease in the number of indications observed during the third ISi was consistent for the other

• (non-beltline) welds examined. The reduction in the number of indications identified (considered to be false positives) is attributed to improved examination detection and analysis techniques used for the third interval. Regardless of which set of indications is considered, every recorded indication was subsurface and acceptable per ASME Code criteria.

The results for the first interval were performed from the inside surface of the RPV and utilized an immersion technique. Because the different techniques utilized between the three reactor vessel examinations prohibits a direct correlation between the recorded indications, the licensee did not attempt to compare the indications found during first interval (immersion technique) to those identified in the second and third (contact technique) intervals.

As discussed above, the NRC staff reviewed the RAI response and 'concludes that the size and locations of the indications found during the second and third intervals are similar, and therefore, it can be concluded that the indications are not growing between the two sets of exams. The differences are attributed to the improved sizing techniques and number of scans/scan increments utilized in the technique qualified under ASME Section XI, Appendix VIII. Thus, the information cited in the submittal and the RAI response, resolves the NRC staffs questions.

Finally, the NRC staff notes that the next inspection for FNP-Unit 1 would be conducted before the end of November 2027. The NRC staff has reviewed the revised PWROG plan and finds thatthe proposed alternative match the inspection plan for FNP-Unit 1.

Based on the above, the NRC staff concludes that the licensee has addressed Plant-Specific Information 3 satisfactorily because the licensee demonstrated that the plant-specific flaw information for FNP-Unit 1 in FNP-ISl-ALT-18, Version 1, is bounded by the WCAP-A, supporting the plant-specific applicability of the WCAP-A to the FNP-Unit 1.

The NRC staff verified the input data and performed independent calculations to check the output results in Table 3 of the alternative. The difference between the licensee's and staff's calculated TWCF9s-TOTAL is insignificant. With this information, the NRC staff concludes that the TWCF9s-TOTAL value in Table 3 of the alternative is bounded by the WCAP-A results.

Consequently, the licensee has demonstrated that the proposed alternative will provide an acceptable level of quality and safety and meets the guidance provided by RG 1.174, Rev. 1, for risk-informed decisions.

The NRC staff considers that the request meets all the conditions and limitations described in the WCAP-A. This conclusion is based on the plant-specific information provided by the licensee bounding the data in the WCAP-A and the discussion and !:?Ubsequent disposition of the one large indication in a beltline plate that was not bounded by Table 3 in 10 CFR 50.61a.

Given the multiple reasons listed in the alternative and the RAI response, the NRC staff concludes that extendng the ISi interval for Categories B-A and B-D components from 1Oto 20 years will result in no appreciable increase in risk. Therefore, FNP-ISl-ALT-18, Version 1, provides an acceptable level of quality and safety, and the alternative can be authorized for Categories B-A and B-D components pursuant to 10 CFR 50.55a(z)(1 ). Further, the NRC staff

_ accepts the alternative examination date (before the end of November 2027) for Categories B-A and B-D components at FNP-Unit 1.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the licensee's proposed alternative provides an acceptable level of quality and safety. 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 use of the licensee's proposed alternative at Farley, Unit 1, for the duration up to November 2027.

All other ASME Code,Section XI, requirements for which relief was not speeifically requested and authorized herein by the staff remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor: Patrick Purtscher, NRR/DE/EVIB Date of issuance: February 1, 2016

ML16013A348 *via e~mail OFFICE LPLll-1/PM LPLll-1/LA DE/EVIB/BC LPLll-1/BC LPLll-1/PM NAME SWilliams SFigueroa JMcHale* MMarkley SWilliams DATE 01/21/16 01/19/16 01/13/16 02/01/16 02/01/16