ML033280602

Prairie, Units 1 and 2, Relief, No. 15, Evaluation of Relief Request No. 15 for the Third 10-Year Interval Inservice Inspection Program

ML033280602
Person / Time
Site:	Prairie Island
Issue date:	01/13/2004
From:	Raghavan L NRC/NRR/DLPM/LPD3
To:	Solymossy J Nuclear Management Co
References
TAC MB8027, TAC MB8028
Download: ML033280602 (6)

Text

January 13, 2004 Mr. Joseph M. Solymossy Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -

EVALUATION OF RELIEF REQUEST NO. 15 FOR THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM (TAC NOS. MB8027 AND MB8028)

Dear Mr. Solymossy:

By letter dated March 14, 2003, the Nuclear Management Company, LLC (NMC), submitted Relief Request No. 15 (RR-15) for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, respectively. NMC revised RR-15 by letter dated November 21, 2003, in response to a Request for Additional Information from the U. S. Nuclear Regulatory Commission (NRC). In RR-15, NMC proposed an alternative to use VT-1 visual examinations in lieu of 100 percent surface examination of all reactor pressure vessel (RPV) closure head nuts. Full surface examination of all RPV closure head nuts is required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, 1989 Edition with no Addenda, Examination Category B-G-1. This is the Code of record during the PINGP, Units 1 and 2, third 10-year inservice inspection (ISI) intervals.

The enclosure provides the NRC staffs safety evaluation (SE) for RR-15. As noted in the SE, the NRC staff concludes that NMCs proposed alternative provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes NMCs proposed alternative, described in the revised RR-15, for the third 10-year ISI intervals.

Sincerely,

/RA/

L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosure:

Safety Evaluation cc w/encl: See next page

ML033280602 *Provided concurrence with comments OFFICE PDIII-1/PM PDIII-1/LA EMCB/SC* OGC** PDIII-1/SC NAME AMcMurtray RBouling SCoffin RHoefling LRaghavan DATE 12/29/03 01/13/04 12/29/03 01/08/04 01/08/04 Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:

Jonathan Rogoff, Esquire Tribal Council Vice President, Counsel & Secretary Prairie Island Indian Community Nuclear Management Company, LLC ATTN: Environmental Department 700 First Street 5636 Sturgeon Lake Road Hudson, WI 54016 Welch, MN 55089 Manager, Regulatory Affairs Nuclear Asset Manager Prairie Island Nuclear Generating Plant Xcel Energy, Inc.

Nuclear Management Company, LLC 414 Nicollet Mall, R.S. 8 1717 Wakonade Drive East Minneapolis, MN 55401 Welch, MN 55089 John Paul Cowan Manager - Environmental Protection Division Executive Vice President & Chief Nuclear Minnesota Attorney Generals Office Officer 445 Minnesota St., Suite 900 Nuclear Management Company, LLC St. Paul, MN 55101-2127 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office Craig G. Anderson 1719 Wakonade Drive East Senior Vice President, Group Operations Welch, MN 55089-9642 Nuclear Management Company, LLC 700 First Street Regional Administrator, Region III Hudson, WI 54016 U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 121 Seventh Place East Suite 200 St. Paul, MN 55101-2145 November 2003

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION RELIEF REQUEST NO. 15 (UNITS 1 AND 2)

NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated March 14, 2003, the licensee, Nuclear Management Company, LLC (NMC),

submitted Request for Relief No. 15 that proposes an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, for Prairie Island Nuclear Generating Plant, Units 1 and 2. In response to a Request for Additional Information from the U.S. Nuclear Regulatory Commission (NRC), the licensee revised the request and provided further information in a letter dated November 21, 2003. The request for relief is for the third 10-year inservice inspection (ISI) interval at Prairie Island Units 1 and 2.

Pacific Northwest National Laboratory, an NRC contractor, has evaluated the revised request for relief and supporting information submitted by the licensee. The NRC staff incorporated the conclusions of the evaluation from Pacific Northwest National Laboratory into this safety evaluation.

2.0 REGULATORY EVALUATION

ISI of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME B&PV Code, and applicable addenda, as required by 10 CFR 50.55a(g),

except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulation at 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the 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.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was

incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of record for the Prairie Island Units 1 and 2 third 10-year interval ISI program, which began, respectively, on December 17, 1993, and December 21, 1994, is the 1989 Edition of Section XI of the ASME B&PV Code, with no addenda.

3.0 TECHNICAL EVALUATION

The information provide by NMC in support of the request for relief from Code requirements has been evaluated and the bases for disposition are documented below:

Request for Relief 15, Examination Category B-G-1, Item B6.10, Reactor Pressure Vessel Closure Head Nuts Code Requirement: Examination Category B-G-1, Item B6.10, requires 100 percent surface examination of all reactor pressure vessel (RPV) closure head nuts.

Licensees Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to perform a visual VT-1 examination in lieu of surface examination as an alternative to the requirements of the Code for the RPV closure head nuts.

Licensees Basis for Proposed Alternative (as stated):

The primary degradation mechanisms leading to failure of nuts are corrosion, cracking, wear, and thread damage. These degradation mechanisms tend to initiate on the surface of the nut and, therefore, surface examination is required by the ASME Code,Section XI. However, detection of degradation can also be made by VT-1 visual examination. Use of the VT-1 visual examinations of the reactor vessel closure head nuts as an alternative to performing the surface examination will not reduce quality or the margin of safety. Within the industry, there have been no failures of RPV closure nuts.

Furthermore, the ASME Subcommittee XI determined that, for the intended purpose of the reactor vessel closure head nuts exams, a VT-1 examination could replace the surface examination. This change in examination method was incorporated into the 1989 Addenda of ASME XI and has remained through the current Edition and Addenda of Section XI. Within 10 CFR 50.55a(b)(2), the NRC has endorsed the use of ASME Section XI through the 1998 Edition with the 2000 Addenda without any limitations on the use of the Table IWB-2500-1, Category B-G-1 requirements. The proposed alternative provides a comprehensive assessment of the condition of the reactor vessel closure head nuts without the need for continual cleaning, re-examination, and handling required by the surface examination method.

NRC Staff Evaluation

The 1989 Edition of ASME Section XI requires that surface examinations of all RPV closure head nuts be performed. These examinations may be performed in-place, under tension, or when the connection is disassembled; however, it is not standard practice to inspect the closure head nuts, in place, under tension, since these components are disassembled each refueling outage. As an alternative to the Code requirements, the licensee proposed to perform a direct VT-1 visual examination in lieu of the Code-required surface examination of the RPV closure

head nuts. The licensee further proposed to use the acceptance criteria of Paragraph IWB-3517.1 of the 1989 Edition of Section XI, which includes requirements for evaluation of crack-like indications and other relevant conditions requiring corrective action, such as deformed or sheared threads, localized corrosion, deformation of part and other degradation mechanisms.

ASME Section XI, Task Group for ISI Optimization performed a survey on RPV bolting which revealed that no service-induced cracking of closure head nuts had occurred in U.S. operating plants. This survey was then used as part of the technical basis for changing the Code-required examination methods from surface to visual examination for Category B-G-1, Item B6.10. The 1989 Addenda and subsequent editions of ASME Section XI have incorporated this visual VT-1 examination requirement in lieu of surface examinations. Since these are later versions of the Code than the current Code of record for the Prairie Island Units 1 and 2 third 10-year interval ISI program, the licensee proposed an alternative to use VT-1 visual examinations.

Typical relevant conditions that would require corrective action prior to putting closure head nuts back into service would include corrosion, deformed or sheared threads, deformation, and degradation mechanisms (i.e., boric acid attack). Surface examination procedures are qualified for the detection of linear type flaws (cracks) with corresponding acceptance criteria for rejectable linear flaw lengths only. When performing surface examinations in accordance with the 1989 Edition of the Code, Item B6.10, the surface examination acceptance criteria is not provided, as it was in the course of preparation. Without clearly defined acceptance criteria, relevant conditions that require corrective measures may not be adequately addressed.

The 1989 Addenda of Section XI, Article IWB-3000, Acceptance Standards, Paragraph IWB-3517.1, Visual Examination, VT-1, describes relevant conditions that require corrective action prior to continued service of bolting and associated nuts. Included for corrective action in IWB-3517.1 is the requirement to compare crack-like flaws to the flaw standards of IWB-3515 for acceptance. Surface examination acceptance criteria are typically limited to linear type flaws (i.e. cracking, aligned pitting and corrosion). Because the VT-1 visual examination acceptance criteria include the requirement for evaluation of crack-like indications and other relevant conditions requiring corrective action such as deformed or sheared threads, localized corrosion, deformation of part, and other degradation mechanisms, it is concluded that the VT-1 visual examination includes a more comprehensive assessment of the condition of the closure head nuts.

4.0 CONCLUSION

The NRC staff has reviewed the licensees submittal and concludes that the alternative VT-1 visual examination proposed by the licensee, in combination with the acceptance criteria in Paragraph IWB-3517.1, provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternative to use VT-1 visual examinations, in lieu of 100 percent surface examination, of all RPV closure head nuts, for the remainder of the third 10-year ISI interval at Prairie Island Units 1 and 2. All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: T. McLellan Date: January 13, 2004