ML20217P828

Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i)

ML20217P828
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	04/07/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217P811	List: ML20217P807 ML20217P828
References
NUDOCS 9804100205
Download: ML20217P828 (3)

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p- t UNITED STATES E

fE NUCLEAR REGULATORY COMMISSION i WASHINGTON, D.C. 2006H001

• • • • • f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR PEGULA RELATED TO REQUEST FOR RELIEF NO.lSl2-08 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNITS 1 AND 2. GRAND GULF NUCLEAR ST RIVER BEND STATION. AND WATERFORD STEAM ELECTRIC STATION.

DOCKET NUMBERS: 50-313. 50-368. 50-416. 50-458. 50-382

1. INTRODUCTION  !

The Technical Specifications for Arkansas Nuclear One, Units 1 and 2, Grand Gulf Nuclear Station, River Bend Station, and Waterford Steam Electric Station, Unit 3 (Entergy sites), sta that the inservice inspection and testing 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 Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(l). The 10 CFR 50.55a(a)(3) states that alternatives 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 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 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 ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated reference in 10 CFR 50.55a(b) on the date twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of the ASME Code,Section XI, for the Entergy sites, is the 1992 edition and portions of the 1993 Addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

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2 Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examina-tion requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support 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(g)(6)(l), the Commission may grant relief and may impose attemative require- '

monts 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.

By letter dated January 29,1998 Entergy Operations, Inc., the licensee for the Entergy sites, requested relief from the requirements of the 1992 Edition,1993 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, in regard to corrective measures for system pressure i tests, as stated in Subsection IWA 5250 (a)(2).

The staff has reviewed and evaluated the licensee's request and the supporting infcrmation on the proposed relief request 1S12-08 for the Entergy sites pursuant to the provisions of 10 CFR l

50.55a(s)(3)(i).

2.0 EVALUATION

Code Reauirement: ASME Section XI,1992 Edition,1993 Addenda, Subartic'e IWA-5250(a)(2) states that if leakage occurs at a bolted connection during a system test, one bolt closest to the  !

leak shall be removed, VT-3 examined, and evaluated for degradation in accordance with IWA-3100.

Licensee's Code Relief.. Request Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requests altomative relief to perform the actions specified in IWA-5250(a)(2) on bolts if leakage occurs at a bolted connection only if a systematic evaluation concludes that a bolt removalis warranted.

Licensee's Basis for Reauestino Relief: Leaking conditions at a bolted connection may be an important variable in the degradation of fasteners. However, leakage is not the only variable, and in some cases may not be the degradation mechanism. Other variables or factors to be )

j considered are: Joint botting materials, service age of joint bolting materials, location of the leakage, history of leakage at the joint, evidence of corrosion with the joint assembled, '

corrosiveness of process fluid, and plant / industry studies of similar botting materials in a similar j

environment. These variables are important to consider before disassembling a bolted '

connection for a visual VT-3 examination. Removal of botting at a mechanical connection may not be the most prudent decision. Entergy considers the requirement to be unnecessarily prescriptive and restrictive.

Licensee's Proposed Altemative Examination: (as stated)

"When leakage is identified at bolted connections by visual VT 2 examination during system pressure testing, an evaluation will be performed. The evaluation will determine the susceptibility  ;

of the botting to corrosioa, assess the potential for failure, and identify appropriate corrective  ;

actions. The following factors will be considered, as necessary, when evaluating the leakage:

1) Bolting materials

2) Corrosiveness of the process fluid 4

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3) Leakage location

4) Leakage history at connection

5) Visual evidence of corrosion at connection (connection assembled)

6) Industry studies and history of similar bolting in similar environment

7) Condition and leakage histoty of adjacent components Furthermore if the initial evaluation indicates the need for a more in-depth evaluation, the actions specified in IWA-5250(a)(2) shall be performed."

Evaluation /

Conclusions:

in accordance with the 1992 edition of the ASME Code,Section XI, in regard to corrective measures for system pressure tests, as stated in the 1993

, Addends, Subsection IWA-5250(a)(2), when leakage occurs at bolted connections, one bolt closest to the leak is required to be removed for VT-3 visual examination, in lieu of the code-required removal of botting to perform a VT-3 visual examination, the licensee has proposed to perform an evaluation of the bolted connection to determine the susceptibility of the bolting to corrosion and the potential for failure. If the initial evaluation indicates the need for a more in-depth evaluation, the actions specified in lWA-5250(a)(2) snall be performed, i.e., the bolt closest to the source of leakage will be removed, VT-3 examined, and evaluated in accordance with IWA-3100(a). This altomative allows the licensee to utilize a systematic approach and sound .

engineering judgement, provided that as a minimum, all of the seven evaluation factors listed in the licensee's proposed altemative are considered. As a result, the licensee's attemative to the code-required removal of bolting at a joint when leakage occurs will provide an acceptable level of quality and safety, as the integrity of the joint will be maintained. Therefore, the staff concludes that the licensee's proposed altemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Principal Contributors: P.' Patnaik D. Wigginton Date: April 7, 1998 l

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