ML15342A043

Supplemental Safety Evaluation to Relief Request I3R-15 for Reactor Vessel Closure Head Penetration Nozzle Repair Technique, Inservice Inspection Program - Third 10-Year Interval

ML15342A043
Person / Time
Site:	Harris
Issue date:	01/06/2016
From:	Benjamin Beasley Plant Licensing Branch II
To:	Waldrep B Duke Energy Progress
Barillas M
References
CAC MF6053
Download: ML15342A043 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 6, 2016 Mr. Benjamin C. Waldrep Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Rd.

M/C HNP01 New Hill, NC 27562-0165

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - SUPPLEMENTAL SAFETY EVALUATION TO RELIEF REQUEST 13R-15 REGARDING RELIEF REQUESTS 13R-09, 13R-11, AND 13R-13 FOR REACTOR VESSEL CLOSURE HEAD PENETRATION NOZZLE REPAIR TECHNIQUE, INSERVICE INSPECTION PROGRAM - THIRD 10-YEAR INTERVAL (CAC NO. MF6053)

Dear Mr. Waldrep:

On May 4, 2015, the U.S. Nuclear Regulatory Commission (NRC) staff verbally authorized the use of Relief Request (RR) 13R-15 for the repair of degraded reactor vessel closure head (RVCH) nozzles 14, 18, and 23 at the Shearon Harris Nuclear Power Plant, Unit 1. The script for the verbal authorization was issued on May 7, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15126A542). By letter dated September 18, 2015, the NRC staff issued a safety evaluation (SE) (ADAMS Accession No. ML15203A702) documenting the staffs detailed technical basis for the verbal authorization, concluding that the licensee adequately addressed all of the regulatory requirements set forth in Title 10 of the Code of Federal Regulations (1 O CFR) 50.55a(z)(1 ).

The purpose of the enclosed supplemental SE is to address the calculation error identified during the review of RR 13R-15. The calculation was previously submitted to support RRs 13R-09, 13R-11, and 13R-13. The NRC authorized RRs 13R-09, 13R-11, and 13R-13 by letters dated October 2, 2012; September 13, 2013; and April 4, 2014 (ADAMS Accession Nos. ML12270A258, ML13238A154, and ML14093A075, respectively). In the enclosed supplemental SE, the NRC staff finds that the error identified in calculation 32-9176350-001, for the heat flow calculations, will not exceed the required interpass temperature limits. The NRC staff finds that the error in calculation 32-9176350-001 did not affect the conclusion in the SEs for RRs 13R-09, 13R-11, and 13R-13, and the subject repairs for RVCH nozzles stated above continue to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

The RRs 13R-09, 13R-11, 13R-13, and 13R-15 apply to the third 10-year inservice inspection interval, which is scheduled to end on May 1, 2017.

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

Sincerely, Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosure:

Supplemental Safety Evaluation cc w/enclosure: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SUPPLEMENTAL SAFETY EVALUATION REGARDING RELIEF REQUESTS 13R-09, 13R-11. AND 13R-13 ALTERNATIVE REPAIR OF REACTOR VESSEL CLOSURE HEAD PENETRATION NOZZLES DUKE ENERGY PROGRESS. INC.

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NUMBER 50-400

1.0 INTRODUCTION

By letters dated October 2, 2012; September 13, 2013; and April 11, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12270A258, ML13238A154, and ML14093A075, respectively), the U.S. Regulatory Commission (NRC) staff authorized Relief Requests (RRs) 13R-09, 13R-11, and 13R-13 for repairs to various degraded reactor vessel closure head (RVCH) penetration nozzles at the Shearon Harris Nuclear Power Plant, Unit 1 (Harris). In RR 13R-09, the NRC authorized repair of nozzles 5, 17, 38, and 63. In RR 13R-11, the NRC authorized repair of nozzle 49; in RR 13R-13, the NRC authorized repair of nozzle 37.

In a letter dated April 29, 2015 (ADAMS Accession No. ML15120A406), the licensee notified the NRC that its evaluation of the impact of a residual flaw in the J-groove weld contained an error as documented in the calculation 32-9176350-001, Shearon Harris Unit 1 CRDM/CET Nozzle As-Left J-groove Weld Analysis. The calculation error affects all nozzles repaired as part of RRs 13R-09, 13R-11, and 13R-13. The licensee previously submitted calculation 32-9176350-001 to support RR 13R-13 (ADAMS Accession No. ML13330A996). The licensee referenced calculation 32-9176350-001 in the original RR 13R-15 submittal dated April 2, 2015 (ADAMS Accession No. ML15092A236). Subsequently, the licensee corrected the original calculation and submitted the revised calculation 32-9176350-002 as documented in Enclosure 3 of the submittal dated April 29, 2015 (ADAMS Accession No. ML15120A406 (non-proprietary version)).

By letter dated September 18, 2015, the NRG issued the safety evaluation (SE) (ADAMS Accession No. ML15203A702) following the verbal authorization granted on May 4, 2015, and documented via script dated May 7, 2015 (ADAMS Accession No. ML15126A542), documenting the staff's technical basis for the verbal authorization of RR 13R-15 for the repair of degraded RVCH nozzles 14, 18, and 23 at Harris.

Enclosure

The purpose of this supplemental SE is to address the calculation error in previously authorized RRs 13R-09, 13R-11, and 13R-13. The NRC staff finds that the error identified in calculation 32-9176350-001, for proposed heat flow calculations, will not exceed the required interpass temperature limits.

The NRC staff finds that the error in calculation 32-9176350-001 did not have a significant impact, because the subject repairs for RVCH nozzles stated above did not significantly reduce the level of quality and safety in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ).

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(z)(1 ), the licensee requested authorization of alternatives to various requirements in IWA-3000, IWA-4000, and IWB-3000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and ASME Code Case N-638-1, "Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW [Gas Tungsten Arc Welding] Temper Bead Technique,Section XI, Division 1."

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI.

Pursuant to 10 CFR 50.55a(g)(6)(ii), 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.

As stated in 10 CFR 50.55a(g)(6)(ii)(D), "Reactor vessel head inspections," licensees of pressurized-water reactors (PWRs) are required to augment their ISi of the RVCH 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," with conditions.

ASME Code Case N-638-1 provides requirements for automatic or machine GTAW of Class 1 components without the use of preheat or postweld heat treatment.

As stated, in part, in 10 CFR 50.55a(z), 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 alternative provides 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 alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

The licensee's original calculation 32-9176350-001 contains an error, because the calculation did not consider the impact of the base metal of the reactor vessel head surrounding the repaired RVCH nozzle on the flaw growth calculation. In addition, the revised calculation 32-9176350-002 calculated the RVCH service life considering all the repaired RVCH nozzle configurations that have been performed to date from 2012 to the spring 2015 refueling outage.

The licensee performed the fracture mechanics analysis to determine the acceptability of leaving degraded J-groove weld and butter material in the reactor vessel head following the repair of either an RVCH nozzle or Core Exit Thermocouple nozzle by the inside diameter temper bead weld procedure. The licensee postulated that a flaw in the RVCH would combine with a large stress corrosion crack in the remnant J-groove weld and butter to form a radial corner flaw that would propagate into the low alloy steel RVCH by fatigue crack growth under cyclic loading conditions.

The licensee used the screening criteria of the ASME Code,Section XI, Appendix C to determine the failure mode and appropriate method of analysis (linear elastic fracture mechanics, elastic-plastic fracture mechanics, or limit load) for flaws in ferritic materials of the RVCH, considering the applied stress, temperature, and material toughness. The licensee used the linear elastic fracture mechanics method to derive the stress intensity factor for the final flaw size to satisfy the available fracture toughness of the RVCH. For the more ductile material, the licensee used the elastic-plastic fracture mechanics method to evaluate flaw stability and crack driving force in the RVCH. The licensee also performed a limit load analysis to ensure that the postulated flaw in the RVCH satisfies the primary stress limits of the ASME Code, Section Ill.

The flaw evaluation using the linear elastic fracture mechanics method and elastic plastic fracture mechanics method did not change significantly between the original calculation 32-9176350-001 and revised calculation 32-9176350-002. The major change involved the limit load analysis.

In the revised calculation 32-9176350-002, the licensee's limit load analysis was based on primary stress limits of the ASME Code, Section 111, NB-3332. As shown in Figures C-1 to C-11 of calculation 32-9176350-002, the licensee modeled the details of the region that surrounds the repaired CROM nozzle bore. The approach of calculating available years of service is based on determining the available area of reinforcement as required per the ASME Code, Section Ill, NB-3332.

The nozzle repair results in removal of the structural material in the RVCH. In addition, the remnant J-groove weld is not considered as structural material as it contains flaws. The licensee also discounted additional area caused by postulated crack growth into the carbon steel of the RVCH from the available structural area. As such, the licensee's nozzle model considered the available reinforcement area surrounding the nozzle, the removed bore area, the area partially removed caused by strength ratio, the removed J-groove weld area caused by crack growth, the removed J-groove weld and buttering area, and the reduced new weld area of reinforcement caused by strength ratio.

The licensee's limit load analysis demonstrated that the as-repaired RVCH continues to satisfy the primary stress limits of NB-3000, considering postulated flaws emanating from the original J-groove weld. This is accomplished by comparing the available reinforcement areas in the vicinity of the repaired nozzles with the areas removed from consideration of carrying primary load, in accordance with NB-3332.

The results of the licensee's revised limit load analysis showed that among the nine nozzle repairs (numbers 5, 14, 17, 18, 23, 37, 38, 49, and 63), RVCH nozzle number 14 yields the most limiting service life of 15 years. The licensee noted that as the earliest repairs were performed in May 2012, the RVCH nozzles are acceptable for 12 years of additional operation, starting from April 2015.

The licensee also noted that based on a combination of linear elastic and elastic-plastic fracture mechanics analysis of a postulated remaining flaw in the original Alloy 182 J-groove weld and butter material, a Harris CROM or CET nozzle is considered to be acceptable for 30 years of operation following the nozzle repair.

For RRs 13R-09, 13R-11, 13R-13, and 13R-15, the NRC staff approved the use of the RR for the third 10-year ISi interval, which commenced on May 2, 2007, and will end on May 1, 2017.

In the NRC's evaluation of RR 13R-9, dated October 2012, the NRC staff noted that the RVCH repaired nozzles are acceptable for the design life of the repair, which is 14.8 effective full power years.

The NRC staff recognizes that the linear elastic fracture mechanics and elastic-plastic fracture mechanics method predicted a service life of 30 years for the RVCH following the nozzle repairs. However, based on the licensee's latest limit load analysis in calculation 32-9176350-002, the service life for the RVCH following the nozzle repairs should be limited to 15 years, based on nozzle 14 being the most limiting case.

The NRC staff notes that since the discovery of indications in the RVCH nozzles in 2012, the licensee has performed, and is required to perform, bare metal visual examinations of every RVCH nozzle every subsequent refueling outage at Harris in accordance with 10 CFR 50.55a(g)(6)(ii)(D). This inspection frequency will adequately monitor the RVCH nozzles to ensure that the repaired nozzles are structurally sound. The inspection frequency also provides a means of verifying the flaw growth calculations to ensure that the projected flaw growth in the remnant J-groove weld will not affect the structural integrity of the RVCH.

The NRC staff finds that as indicated by the results of the licensee's revised calculation for the remnant J-groove weld, the error in the licensee's original calculation did not affect the structural integrity of the RVCH. Therefore, the NRC staff determines that the error in the original calculation did not affect the safety function of the RVCH.

4.0 CONCLUSION

Based on the information submitted, the NRC staff concludes that the error in the licensee's calculation 32-9176350-001 did not affect the structural integrity of the repaired RVCH nozzles nor the RVCH itself. Accordingly, the NRC staff concludes that the repaired nozzles remain in compliance with the requirements of the ASME Code, Sections Ill and XI, Code Case N-638-1 and Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D), for which relief was not requested. Therefore, the NRC staff concludes that the licensee's calculation error did not affect the conclusion in the SE for RRs 13R-09, 13R-11, and 13R-13, and the repairs continue to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1 ).

Principal Contributor: John Tsao Date:January 6, 2016

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

Sincerely,

/RAJ Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosure:

Supplemental Safety Evaluation cc w/enclosure: Distribution via Listserv DISTRIBUTION:

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