ML14353A324

Relief from the Requirements of the ASME Code,Section XI

ML14353A324
Person / Time
Site:	Harris
Issue date:	01/05/2015
From:	Lisa Regner Plant Licensing Branch II
To:	Waldrep B Carolina Power & Light Co
Barillas M DORL/LPL2-2 301-415-2760
References
TAC MF4113
Download: ML14353A324 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Benjamin C. Waldrep Vice President Shearon Harris Nuclear Power Plant Duke Energy Progress, Inc.

5413 Shearon Harris Rd.

New Hill, NC 27562-0165 January 5, 2015

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE, SECTION XI (TAC NO. MF4113)

Dear Mr. Waldrep:

By letter dated April 25, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14115A419), as supplemented by letter dated September 8, 2014 (ADAMS Accession No. ML14251A274), Duke Energy Progress, Inc. (the licensee), submitted a relief request to the U.S. Nuclear Regulatory Commission (NRC) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP-1 ). The proposed relief request would defer the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI required volumetric examination of the HNP reactor vessel full penetration pressure retaining Categories B-A and B-D welds for the third inservice inspection (lSI) interval, currently scheduled for 2015, and perform the required examinations in 2024, plus or minus one refueling outage.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 55a(a)(3)(i),

Relief Request 13R-14 contains a proposed alternative to extend the lSI interval for examinations of the reactor pressure vessel welds (Category B-A), as well as the nozzle-to-vessel welds and inner radius sections (Category B-D), from 10 years to 20 years. 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 alternative for Categories B-A and B-D components examinations pursuant to 10 CFR 50.55a(a)(3)(i) at HNP-1 until the end of the third interval, which is now 2027. for the subject components.

Further, the NRC staff authorizes the alternative examination date for Categories B-A and B-D components for HNP Units in 2024, plus or minus one refueling cycle.

The current third 10-year interval commenced on May 2, 2007, and ends on May 1, 2017.

All other ASME Operation and Maintenance Code requirements for which alternatives or relief were not specifically requested and approved in the subject request remain applicable.

If you have any questions, please contact the Project Manager, Martha Barillas, at 301-415-2760, or by e-mail at Martha.Barillas@nrc.gov.

Docket No. 50-400

Enclosure:

Safety Evaluation cc w/enclosure: Distribution via ListServ Lisa M. Regner, Acting 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 REQUEST FOR RELIEF 13R-14 REGARDING THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL DUKE ENERGY PROGRESS. INC SHEARON HARRIS NUCLEAR POWER PLANT. UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By letter dated April 25, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14115A419), as supplemented by letter dated September 8, 2014 (ADAMS Accession No. ML14251A274), Duke Energy Progress, Inc. (the licensee), requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, as currently implemented in the inservice inspection (lSI) program for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP-1) lSI program, third 1 0-year interval.

Specifically, Relief Request (RR) 13R-14 contains a proposed alternative pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR) to extend the lSI interval for examinations of the reactor pressure vessel (RPV) welds (Category B-A), as well as the nozzle-to-vessel welds and inner radius sections (Category B-D) from 10 to 20 years.

The current third 10-year interval commenced on May 2, 2007, and ends on May 1, 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 lSI of ASME Code Class 1, 2, and 3 components and system pressure tests during the first 1 0-year interval and subsequent 1 0-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(b),

subject to the limitations and modifications listed therein. For the third 1 0-year lSI interval at HNP, 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, "Rules for lnservice Inspection of Nuclear Power Plant Components."

Enclosure Title 10 CFR 50.55a(a)(3) 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). Per 10 CFR 50.55a(a)(3)(i), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety.

Regulatory Guide (RG) 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"

describes methods 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 1.17 4, 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 U.S. Nuclear Regulatory Commission (NRC or the Commission), 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.

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 alternative requested by the licensee.

2.2 Background

The lSI of Categories B-A and B-D components consists of 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 in the ASME Code,Section XI.

2.3 Summary ofWCAP-16168-NP-A. Rev. 2 In June 2008, the Pressurized Water Reactor Owners Group (PWROG) issued the NRC-approved topical report WCAP-16168-NP-A, Rev. 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (ADAMS Accession No. ML082820046), which is in support of a risk-informed assessment of extensions to the lSI intervals for Categories B-A and B-D components. Specifically, WCAP-16168-NP-A, Rev. 2, took data associated with three different pressurized-water reactor (PWR) plants (referred to as the pilot plants), one designed by each of the three main vendors (Westinghouse, Combustion Engineering, and Babcock and Wilcox (B&W)) for PWR nuclear power plants in the United States, and performed studies on these pilot plants to justify the proposed extension of the lSI interval for Categories B-A and B-D components from 10 to 20 years.

The analyses in WCAP-16168-NP-A, Rev. 2, used probabilistic fracture mechanics (PFM) tools and inputs from the work described in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61):

Summary Report" (ADAMS Accession No. ML061580318) and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (ADAMS Accession No. ML070860156). The PWROG analyses incorporated the effects of fatigue crack growth and lSI.

Design basis transient data was used as input to the fatigue crack growth evaluation. The effects of lSI were modeled consistent with a previously approved PFM Code in WCAP-14572-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection Topical Report" (ADAMS Accession Nos. ML012630327, ML012630349, and ML012630313). These effects were considered in the PFM evaluations using the Fracture Analysis of Vessels: Oak Ridge computer code (ADAMS Accession No. ML042960391 ). All other inputs were identical to those used in the PTS risk re-evaluation underlying 10 CFR 50.61a.

From the results of the studies, the PWROG concluded that the ASME Code,Section XI 1 0-year inspection interval for Categories 8-A and 8-D components in PWR RPVs can be extended to 20 years. The PWROG 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 ofWCAP-16168-NP-A, Rev. 2) are bounded by the pilot plants.

2.4 Summary of the July 26, 2011, NRC Safety Evaluation for WCAP-16168-NP-A Rev. 2 Th'e original safety evaluation (SE) in WCAP-16168-NP-A, Rev. 2 that was published in 2008 was superseded by the July 26, 2011, SE (ADAMS Accession No. ML111600303) to address the PWROG's request for clarification of the information needed in applications utilizing WCAP-16168-NP-A, Rev. 2. The NRC staff's conclusion in this latter SE determined that the methodology presented in WCAP-16168-NP-A, Rev. 2, 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-16168-NP-A, Rev. 2, the licensee's application must provide the following plant-specific information:

(1) Licensees must demonstrate that the embrittlement of their RPV is within the envelope used in the supporting analyses. Licensees must provide the 951h percentile total through-wall cracking frequency (TWCFrorAd and its supporting material properties at the end of the period in which the relief is requested to extend the lSI from 10 to 20 years. The 951h percentile TWCFrorAL 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, b.T3o, must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRC-approved 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 the PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.

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

In its request for an 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-1 0-238 (ADAMS Accession No. ML11153A033).

(4) Licensees with 8&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 bound the fatigue crack growth for all of its design basis transients, and (b) identify the design bases transients that contribute to significant fatigue crack growth.

(5) Licensees with RPVs having forgings that are susceptible to underclad cracking and with RT MAX-Fo values exceeding 240 a Fahrenheit must submit a plant-specific evaluation to extend the inspection interval for ASME Code,Section XI, Categories 8-A and 8-D RPV welds, from 10 to a maximum of 20 years because the analyses performed in WCAP-A are not applicable.

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

WCAP-16168-NP-A, Rev. 3, which contains this latter SE for WCAP-16168-NP-A, Rev. 2, was issued in October 2011 (ADAMS Accession No. ML11306A084).

3.0 PROPOSED ALTERNATIVE FOR SHEARON HARRIS NUCLEAR POWER PLANT 3.1 Description of Proposed Alternative In its submittal, the licensee requested to defer the ASME Code required Categories 8-A and 8-D welds lSI for HNP-1, currently scheduled for 2015, until2024, plus or minus one refueling outage. The licensee stated this schedule is consistent with the latest revised lSI implementation plan documented in PWROG Letter OG-10-238.

3.2 Components for Which Relief is Requested The affected components are the HNP RPV and RPV 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 request:

For HNP RR 13R-14:

Exam Category 8-A 8-A 8-A 8-A 8-A 8-A 8-D 8-D Item No.

81.11 81.12 81.21 81.22 81.30 81.40 83.90 83.100 Description Circumferential Shell Welds Longitudinal Shell Welds Circumferential Head Welds Meridional Shell Welds Shell-to-Flange Weld Head-to-Flange Weld Nozzle-to-Vessel Welds Nozzle Inner Radius Areas 3.3 Basis for Proposed Alternatives The licensee stated that the methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Categories B-A and B-D components is contained in WCAP-16168-NP-A, Rev. 3. This methodology used estimated TWCF as a measure of the risk of RPV failure. It was demonstrated that the inspection interval for the affected components can be extended from 1 0 to 20 years based on this risk, meeting the change in risk guidelines in RG 1.17 4. The licensee addressed the plant-specific information discussed in Section 2.4 of this SEas follows:

(1) A plant-specific analysis with identified critical parameters and detailed TWCF calculations demonstrated that the HNP RPV parameters are bounded by corresponding pilot plant parameters. The total TWCFs were calculated as 6.59 x 10-11 events per year, less than the value of 1. 76 x 1 o-s events per year for the Westinghouse pilot plant in WCAP-16168-NP-A, Rev. 3.

(2) The frequencies of the HNP RPV limiting design basis transients are bounded by the frequencies identified in the PWROG fatigue analysis.

(3) The results of the previous RPV inspections for HNP are provided, which confirm that satisfactory examinations have been performed on the HNP RPV. The RPV examinations currently scheduled for 2015 for HNP will be deferred until 2024, plus or minus one refueling cycle. The dates provided must be within plus or minus one refueling cycle of 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 HNP. Since HNP 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(a)(3)(i), that the NRC authorize the reliefs.

3.4 Duration of Proposed Alternatives This alternative would extend the duration of the third lSI interval for the subject components with the next ASME Categories 8-A and B-D RPV weld inspections scheduled in 2024.

4.0 TECHNICAL EVALUATION

The NRC staff reviewed RR 13R-14 in detail. In Table 1 of the RR, the "Frequency and Severity of Design Transients" of HNP were found to be bounded by WCAP-16168-NP-A, Rev. 3, based on the information in the license renewal application (LRA) that was approved in 2008 for HNP.

This LRA information confirmed that the projected number of the 60-year design basis transients is below the number specified in the 40-year design basis and were bounded by the assumptions in the methodology. Therefore, the NRC staff determined that the licensee has addressed plant-specific Information Item 2 satisfactorily and confirmed that, regarding design transients, WCAP-16168-NP-A, Rev. 3 methodology is applicable to HNP. Also, the HNP RPV has a single-layer cladding on the inside, consistent with the assumption used in WCAP-16168-NP-A, Rev. 3 analysis. Considering these critical plant-specific parameters, the licensee performed TWCF calculations using WCAP-16168-NP-A, Rev. 3 methodology to address plant-specific Information Item 1.

These TWCF calculations used inputs from Table 3 of the RR. The RR used RG 1.99, Rev. 2, Position 1.1 and Position 2.1, when credible surveillance data was available, to calculate ~ T 3o for all RPV beltline materials for HNP. The NRC staff confirmed that the neutron fluence for all RPV materials and the chemistry factors for the RPV materials based on surveillance data were consistent with the values reported in the HNP license renewal and measurement uncertainty power uprate applications (ADAMS Accession Nos. ML082340985 and ML11356A096 respectively) and are, therefore, acceptable. The NRC staff then independently conducted all calculations in Table 3 and confirmed the accuracy of the applicant's values. They determined that the licensee has addressed plant-specific Information Item 1 satisfactorily and confirmed that the embrittlement of the HNP RPVs is within the envelope used in the Westinghouse pilot plant analysis.

Table 2 of the submittal contains additional information pertaining to previous RPV inspections and the schedule for future ones. Table 2 was provided to address plant-specific Information Item 3. Specifically, Table 2 lists two indications that were identified in the most recent lSI of the HNP RPV beltline region. The licensee concluded that the indications are acceptable per Table IWB-351 0-1 of Section XI of the ASME Code. The NRC staff requested further details concerning the indications, and the applicant provided these in its request for additional information (RAI) response. This response contained details concerning the location, nature, and disposition of the indications. The staff reviewed these details and finds the licensee's dispositioning of the indications is consistent with ASME Code,Section XI requirements. The next inspection for HNP will be conducted in 2024, plus or minus one refueling outage. The staff has reviewed the PWROG plan and finds that the proposed alternatives match the inspection plan for the PWR fleet. Based on the above evaluation, the NRC staff concludes that the licensee has addressed plant-specific Information Item 3 satisfactorily, since the licensee demonstrated that the plant-specific flaw information for HNP is bounded by WCAP-16168-NP-A, Rev. 3. The staff also confirmed that plant-specific Information Items 4-6 do not apply to HNP.

In summary, the NRC staff has reviewed the licensee's submittal and the response to the NRC staff's RAI supplementing the RR. In addition, the NRC staff performed independent calculations to verify the input data and output results in Table 3. With this information, the NRC staff concluded that the TWCFss-TOTAL value in Table 3 is bounded by WCAP-16168-NP-A, Rev. 3 results. In addition, the applicant appropriately addressed all applicable plant-specific information items. Consequently, the licensee has demonstrated that the proposed alternatives will provide an acceptable level of quality and safety and meet the guidance provided by RG 1.174, Rev. 1, for risk-informed decisions.

5.0 CONCLUSION

The NRC staff has completed its review of HNP RR 13R-14. The staff concludes that the plant-specific information provided by the licensee is bounded by the data in WCAP-16168-NP-A, Rev. 3, and that the requests met all the conditions and limitations described in WCAP-16168-NP-A, Rev. 3. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i}, is in compliance with ASME Code requirements, and HNP RR 13R-14 provides an acceptable level of quality and safety. The NRC authorizes the alternative for Categories B-A and B-D components pursuant to 10 CFR 50.55a(a)(3)(i) at HNP-1 until the end of the third examination interval, which is now 2027, for the subject components. Further, the NRC staff authorizes the alternative examination date for Categories 8-A and B-D components for HNP Units in 2024, plus or minus one refueling cycle.

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

Principal Contributors: Dan Widrevitz Date: January 5, 2015

ML14353A324 Sincerely, IRA/

Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

• via memo OFFICE N RR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LAiT NRR/DORL/LPL2-2/LA NRR/DE/EVIB/BC*

NAME MBarillas LRonewicz BCiayton SRosenberg DATE 12/31/14 12/30/14 12/30/14 11/04/14 OFFICE NRR/DORL/LPL2-2/BC(A) NRR/DORLILPL2-2/PM NAME LRegner MBarillas DATE 12/31/14 01/05/15