ML20197A753

Safety Evaluation Accepting License Request for Relief from ASME Code,Section IX Requirements Re Inservice Insp of RHR Heat Exchanger Nozzle Welds & Reactor Vessel Closure Head Nuts

ML20197A753
Person / Time
Site:	Seabrook
Issue date:	02/27/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20197A749	List: ML20197A743 ML20197A753
References
NUDOCS 9803090372
Download: ML20197A753 (2)

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UNITED STATES s* j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066H001

. . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 3

OF THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REQUESTS FOR RELIEF IR-6 AND IR-7 wa NORTH ATLANTIC ENERGY SERVICE CORPORATION SEAEROOK STATION. UNIT NO.1 DOCKET NO. 50-443

1.0 INTRODUCTION

i The Technical Specifications (TS) for Seabrook Station, Unit No.1 state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable andenda as required by 10 CFR 50.55a(g), except where specsc written relief has been granted by the Commission purt uant to 10 CFR 50.55a(g)(6)(i).

10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulties 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 t te requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of ciesign, geometry. and materials of construction of the components. The regulations require that ictervice 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 incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the sted of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Seabrook, Unit 1, first 10-year inservice inspection (ISI) interval is the 1983 Edition through Summer 1983 p Addenda.

Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Se: tion XI of the ASME Code is not practical for its f acility, information sha:1 be submitted to the Commission in suppcrt of that determination and a request

' made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(p)(6)(i), the Commisslon may grant relief and may impose hf $b o

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attemative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.

in a let*r dated October 22.1996, North Atlantic Energy Service Corporation (the licensee),

submitted to the NRC its First 10-Year Inservice Inspection Interval Program Plea Requests for Relief IR-6 and IR-7, for Seabrook Station, Unit 1. Additional information was provided by the licensee in its letter dated May 6.1997, 2.0 EVALUATION l

The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its First 10-Year Inservice inspection Interval Program Plan Requests for Relief IR-6 and IR-7 Based on the information submitted, the staff adopts the contractor's conclusions and re:

• mendations presented in the associated INEEL Technical Letter Report (TLR).

Request for Relief RI-6: The Code requires for Table IWC-2500-1, Examinatior, Category C-B, Item C2.31, a 100% surface examination of heat exchanger nozzle-to-shell welds as defined by Figure IWC-2500-4(c).

Item C2.33 requires a VT-2 visual examination for evidence of leakage. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee sposed attematives to the Code required surface and VT-2 visual examination of Class 2, Residual Heat Removal heat exchanger nozzle-to-shell Welds RH E-98-NZA-B and RH E-98-NZB-B. The licensee proposed the following attomatives, as stated:

Perform a volumetric examination of the nozzle to shell welds.

Perform a 100% surface examination of the nozzle to shell welds.

Perform an inservice system leakage test and VT-2 visual examination per Table IWC-2500-1, Category C-H of the nozzle to shell welds.

The Code requires that the subject examination areas be 100% surface examined and VT-2 visually examined at each inspection interval. Because the reinforcement plate is located inside the vessel, the Code does not address the configuration of the examination area. The licensee has proposed to perform a 100% volumetric and surface examination of the subject nozzle-to-vessel welds. In addition, the licensee vill perform a VT-2 visual examination of the subject exambation areas during pressure tests. The staff determined that the proposed attemative exair.inations exceed Code requirements for the subject nozzle-to-vessel welds. Thus, the licensee's proposed attemative provides reasonable assurance of structuralintegrity for nozzle-to-shell Welds RH E-9B-NZA-B and RH E-9B-NZB-B.

Therefore, the licensee's proposed attemative contained in Request for Relief IR-6 is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Request for Relief RI-7: The Code for Section XI, Table IWB-2500-1, Examination Category B-G-1, item B6.10 requires a 100% surface examination of all reactor vessel closure head nuts.

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i Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an attemative to the Code-required surface examination of the reactor vessel closure head nuts as specified in Table IWB 2500-1 of the 1983 Edition through the Summer 1983 Addenda of ASME Section XI. The licensee proposed the following attemt.tive, as stated:

Perform a VT-1 visual examination as pecified in Table IWB-2500-1, item B6.10 of the 1992 Edition of ASME Section XI, and the acceptance criteria of IWB 517.

The licensee has requested relief from performing the Code-required surface examination on the reactor pressure vessel closure head nuts. As an attemative, the licensee proposes to perform a VT-1 visual examination. Based on a review of examination requirements for Examination Category B-G-1, it is noted that a surface examination is required for only the reactor pressure vessel closure head nuts and the closure studs (when removed). Examination requirements for other B-G-1 items include VT-1 visual and volumetric examinations only.

Article IWB-3000, Acceptance Standards, IWB-3617.1, VisualExamination, VT-1, describes l conditions that require corrective action prior to continued service of botting and associated nuts. lWB-3517.1 requires crack-like flaws to be compared to the flaw standards of IWB-3515 for acceptance. Because the VT-1 visual examination acceptance criteria include the requirement for evMuation of crack-like indications and other relevant conditions requiring corrective action, such as deformed or sheared threade, localized corrosion, defo:mation of pa,t, and other degradation mechanisms, it can be concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of tne closure head nut than a surface examination. As a result, the staff determined that VT-1 visual examination provides an acceptable level of quality and safety, in addition, it is noted that the 1989 Addenda and later Editions of Section XI have changed the requirement for the subject reactor pressure vessel closure head nuts to a VT-1 visual examination and provides appropriate acceptance criteria.

Based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the 1989 Addenda and later editions of the Code require only a VT-1 visual examination on reactor pressure vessel closure head nuts, an acceptable level of quality and safety is provided by the licensee's proposed attemative. Therefore, the licensee's proposed alternative VT-1 visual examination is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3. CONCLUSIONS The staff has evaluated and concluded that the licensee's proposed attematives contained in Requests for Relief IR-06 and IR-07 provide an acceptable level of quality and safety.

Therefore, the licensee's pioposed attematives contained in Requests for Relief IR-06 and IR-07 are authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Principa' Contributor. T. McLellan Dated: February 27, 1998

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