ML18283A049

Request for Relief I4R-6 from ASME Code Visual Examination Requirements for Reactor Vessel Head Penetration Nozzle Weld Specified by Code Case N-729-4

ML18283A049
Person / Time
Site:	Wolf Creek
Issue date:	10/18/2018
From:	Robert Pascarelli Plant Licensing Branch IV
To:	Heflin A Wolf Creek
Singal B, NRR/DORL/LPL4-1, 415-3016
References
EPID L-2018-LLR-0068
Download: ML18283A049 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 18, 2018 Mr. Adam C. Heflin President and Chief Executive Officer, Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1 - REQUEST FOR RELIEF 14R-06 FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE VISUAL EXAMINATION REQUIREMENTS FOR REACTOR VESSEL HEAD PENETRATION NOZZLE WELD SPECIFIED BY CODE CASE N-729-4 (EPID L-2018-LLR-0068)

Dear Mr. Heflin:

By letter dated May 2, 2018, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request (RR) 14R-06 to the U.S. Nuclear Regulatory Commission (NRC) to request relief from the visual examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Code Case N-729-4, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," as required and conditioned by Title 10 of the Code of Federal Regulations ( 10 CFR) paragraph 50.55a(g)(6)(ii)(D), at the Wolf Creek Generating Station, Unit 1 (WCGS).

Specifically, pursuant to 10 CFR 50.55a(z)(1 ), the licensee requested to use the proposed alternative in RR 14R-06 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 licensee has demonstrated that the 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 RR 14R-06 at the WCGS for Refueling Outage (RFO) 22. After RFO 22, the licensee will return to the normal inspection protocol for the remainder of the fourth 10-year inservice inspection interval.

On May 10, 2018, the NRC staff verbally authorized the use of RR IR-06 at WCGS, during RFO 22. This letter and the enclosed safety evaluation provide the written followup of the NRC staff's basis for the verbal authorization.

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

A. Heflin If you have any questions concerning this matter, please contact the Project Manager, Mr. Balwant K. Singal at 301 415-3016 or via e-mail at Balwant.Singal@nrc.gov.

Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF 14R-06 FROM VISUAL EXAMINATION REQUIREMENTS OF CODE CASE N-729-4 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482

1.0 INTRODUCTION

By letter dated May 2, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18128A186), Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request (RR) 14R-06 to the U.S. Nuclear Regulatory Commission (NRC) to request relief from the visual examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Code Case N-729-4, "Alternative Examination Requirements for PWR [Pressurized Water Reactor]

Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," as required and conditioned by Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(g)(6)(ii)(D), "Augmented ISi [inservice inspection]

requirements: Reactor vessel head inspections," at the Wolf Creek Generating Station, Unit 1 (WCGS). Specifically, pursuant to 10 CFR 50.55a(z)(1 ), the licensee requested to use the proposed alternative in RR 14R-06 on the basis that the alternative provides an acceptable level of quality and safety.

On May 10, 2018 (ADAMS Accession No. ML18131A183), the NRC verbally authorized the use of RR 14R-06 at WCGS for the alternative examination of reactor vessel closure head (RVCH)

Nozzle No. 1 before return to service in Refueling Outage (RFO) 22.

2.0 REGULATORY EVALUATION

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)( 4 ), "lnservice inspection standards requirement for operating plants," which states, in part, that ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in Section XI of the ASME Code.

Pursuant to 10 CFR 50.55a(g)(6)(ii), "Augmented ISi program," "[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."

Enclosure

The regulation at 10 CFR 50.55a(g)(6)(ii)(D) requires licensees of PWRs to augment their ISi of the reactor vessel head nozzles in accordance with ASME Code Case N-729-4 with conditions as a result of operating experience of primary water stress corrosion cracking in reactor vessel head nozzles.

The regulations at 10 CFR 50 .55a(z) states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The licensee must demonstrate that (1) the proposed alternatives 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 the use of an alternative and the NRC to authorize the proposed alternative.

3.0 TECHNICAL EVALUATION

3.1 ASME Code Components Affected

Component: RVCH Nozzle No. 1 Code Class: Class 1 Examination Category: 8-P Code Item Number: 84.10 (ASME Code Case N-729-4) 3.2 Applicable Code Edition and Addenda The fourth 10-year ISi interval edition for the WCGS is the ASME Code,Section XI, 2007 Edition through 2008 Addenda. The examinations of the RVCH nozzles are performed in accordance with ASME Code Case N-729-4, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D).

3.3 Applicable Code Requirements Table 1 of ASME Code Case N-729-4 requires a bare metal visual examination (VE) to be performed each RFO while implementing Notes 1, 2, 3, and 4 of Table 1.

3.4 Reason for Request During RFO 22, the licensee performed a VE of its RVCH nozzles in accordance with ASME Code Case N-729-4, Table 1, Item No. 84.10. The licensee was able to complete the VEs on all nozzle penetrations except RVCH Nozzle No. 1. The licensee was not able to complete the VE because the annulus region of RVCH Nozzle No. 1 was obscured by a foreign substance.

The licensee was unsuccessful in its attempt to remove the substance from the inspection area of RVCH Nozzle No. 1 and had to resort to more aggressive methods. More aggressive methods of removal would not meet the requirements of N-729-4 VE inspections.

Therefore, the licensee was not able to be perform a VE of RVCH Nozzle No. 1 as required by ASME Code Case N-729-4, Table 1 Item No. 84.10.

3.5 Licensee's Proposed Alternative and Basis for Use In lieu of performing an acceptable VE required by Table 1, Item No. 84.10 of the ASME Code Case N-729-4, the licensee proposed (a) to perform a volumetric examination of essentially 100 percent of RVCH Nozzle No. 1; (b) to perform a demonstrated volumetric leak path assessment of the nozzle penetration; (c) if an unacceptable indication was detected by the proposed demonstrated volumetric leak path assessment or volumetric examination, it would revert to the requirements of ASME Code Case N-729-4 and 10 CFR 50.55a(g)(6)(ii)(D) during RFO 22; (d) to clean the RVCH Nozzle No. 1 penetration and perform a supplemental VE prior to returning to service in RFO 22; and (e) following the return to service, to follow the inspection requirements in accordance with ASME Code Case N-729-4, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D), for the duration of the fourth 10-year ISi interval.

The licensee has previously volumetrically examined all the RVCH nozzle penetrations in 2006, 2013, and 2016. The licensee did not find any degradation during any of these examinations.

The licensee proposed to provide further assurance that no new leak path has been developed by comparing the results from the previous two examinations to the results from the examination performed in RFO 22.

3.6 Duration of Proposed Alternative The licensee stated that the alternative is applicable to RFO 22. After RFO 22, the licensee will return to the normal inspection protocol for the remainder of the fourth 10-year ISi interval.

3. 7 NRC Staff Evaluation The NRC staff determined that while the demonstrated volumetric leak path is not equivalent to a fully-qualified surface leak path assessment, the licensee identified sufficient operational experience and technical basis to show that its proposed alternative provides an acceptable level of quality and safety.

The licensee showed that there has been no previously identified cracking or leakage from the RVCH Nozzle No. 1 penetration and associated J-groove weld at WCGS. The NRC staff noted that while this does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future.

For the technical basis, the licensee compared its previous leak path assessment in RFO 21 performed by Wesdyne International with its current leak path assessment performed by Framatome. Both sets of results show an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7142, "Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation," August 2012 (ADAMS Accession No. ML12241A160), which found, in part, the use of a properly focused zero degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information in NUREG/CR-7142 does demonstrate the effectiveness of the volumetric leak path examination to detect low leakage rates, as performed in accordance with the licensee's proposed alternative.

The NRC staff finds that the volumetric examination of the nozzle and the demonstrated volumetric leak path are NRC approved inspection techniques in accordance with ASME Code Case N-729-4, and that together they provide reasonable assurance of detecting a leak from either the penetration nozzle or the associated attachment weld. The NRC staff finds that the foreign object only obscured the licensee's ability to verify that the penetration was not leaking, and did not prevent the licensee from verifying the general integrity of the low alloy steel head. Therefore, the NRC staff recognizes that the only concern for the missed inspection coverage was the ability to confirm no leakage through the penetration nozzle or weld.

Therefore, to address the inspection limitation imposed by the foreign object at nozzle Penetration No. 1, the NRC staff finds that the use of the volumetric examination of the nozzle and demonstrated volumetric leak path assessment are adequate to provide reasonable assurance of leak tightness and structural integrity of the RVCH.

Therefore, the NRC staff concludes that the licensee's proposed alternative provides reasonable assurance of structural integrity of the RVCH Nozzle No. 1 until the next scheduled examination.

4.0 CONCLUSION

Based on the above, the NRC staff determines that the licensee has demonstrated that the 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 RR 14R-06 at the WCGS for RFO 22. After RFO 22, the licensee will return to the normal inspection protocol for the remainder of the fourth 10-year ISi interval.

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

Principal Contributor: S. Cumblidge, NRR Da~: October 18, 2018

ML18283A049 *Memo dated OFFICE NRR/D0RL/LPL4/PM NRR/D0RL/LPL4/LA NRR/DMLR/MPHB/BC* NRR/D0RL/LPL4/BC NAME BSingal PBlechman SRuffin RPascarelli DATE 10/17/18 10/15/18 9/25/18 10/18/18