ML20140C767

SER Supporting Listed Requests for Relief from ASME Code Section XI Re Inservice Insp Plan Requirements

ML20140C767
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/13/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20140C483	List: ML20140C478 ML20140C767 ML20140C780
References
NUDOCS 8601280399
Download: ML20140C767 (9)

Text

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UNITED STATES p ig NUCLEAR REGULATORY COMMISSION g WASHINGTON, D. C. 20655 v e

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO REQUESTS FOR RELIEF FROM INSERVICE INSPECTION REQUIREMENTS FOR ARKANSAS NUCLEAR ONE, UNIT 1 j

l ARKANSAS POWER AND LIGHT COMPANY DOCKET 50-313 INTRODUCTION Technical Specification 4.2.A for Arkansas Nuclear One, Unit 1, states that inservice examination of 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 Nuclear Regulatory Coninission. Certain requirements of later editions and addenda of Section XI are impractical to perform on older plants because of the plants' design, component geometry, and materials of construction. Thus, 10 CFR 50.55a(g)(6)(1) authorizes the Commission to grant relief from those requirements upon making the necessary findings.

By letters dated August 27, 1984, January 31, 1985, February 4,1985, March 8, 1985 and June 24, 1985, Arkansas Power and light Company submitted its inservice inspection program, revisions, or additional information related to requests for relief from certain Code requirements determined to be impractical to perform on Arkansas Nuclear One, Unit 1, during the inspection interval. The program is based on the requirements of the 1980 Edition through Winter 1981 Addenda of Section XI of the ASME Code, and remains in effect until December 19, 1994 for Unit 1, unless the program is modified or changed prior to the interval end date.

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2 EVALUATION Requests for relief from the requirements of Section XI which have been determined to be impractical to perform have been reviewed by the NRC staff's contractor, Science Applications International Corporation. The con-tractor's evaluations of the licensee's requests for relief and his recommendations are presented in the Technical Evaluation Report (TER)

I attached (Attachment 1). We have reviewed the TER and adopt the evaluations and recommendations. A summary of the determinations made by the staff is presented in the following tables:

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i TABLE 1 '"

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CLASS 1 COMPONENTS i

LICENSEE IWB-2500-1 IWB-2500-1 SYSTEM OR PROPOSED ITEM NO. AREA TO BE REQUIRED ALTERNATIVE EXAM. CAT. COMPONENT RELIEF REQUEST EXAMINED _ METHOD EXAMINATION STATUS I B5.10 B-F Reactor 1 Nom. Pipe Volumetric Vessel Volumetric Granted Size > 4 in. & Surface Exam From Provided Nozzle-to-Safe Vessel ID End Butt Welds Either Pro-(Core Flood posed Pro-i Nozzle-to-Safg cedure is <

End Welds01-025 Demonstrated j

& 01-026) to be Capable j of Detecting

) 0D Flaws or  !

i Surface Exam t is Performed.

i 89.11 B-J Reactor

Nom. Pipe Surface &

Coolant Volumetric Granted

! Size > 4 in. Volumetric Exam From Provided System Circumferential j Vessel ID Either Pro-

! Welds (Hot Leg posed Pro-j Nozzle-to-Safe End Welds01-019 cedure is

{ Demonstrated

& 01-022) to be Capable j

of Detecting j 00 Flaws or

) Surface Exam i

I is Performed.

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. *::SL i IF . iOO-1 IWB-2500-1 SYSTEM OR AREA TO BE PROPOSED -

ITt .40. EXAM. CAT. COMPONENT REQUIRED ALTERNATIVE RELIEF REQUEST EXAMINED METHOD EXAMINATION STATUS B9. B-J Piping Nom. Pipe Surface None, Due Pressure Size < 4 in, ' Granted Pro-Boundary to Weld vided Surface Circumferential Inaccessibility Exam of First Welds (High Pres-sure Injection Accessible

• Pressure Line Welds Boundary Weld 20-008A,21-026, Adjacent to 23-024A,&22-027A, Core Flood Line Inaccessible Welds 19-0208 &

Weld & Visual (VT-2) Exam 19-021A) During System Pressure Tests dre Performed B12.10 B-L-1 Reactor Pump Volumetric Volumetric Relief Not Coolant Casing Pump Exam of Acces- Required Welds sible 95% of Weld B12.20 B-L-2 Pump Pump Casing Visual Partial Granted Pro-(VT-3) Surface vided Code-Replication Required From Volu- Visual. Exam metric Exam is Performed Upon Pump Disassembly for Maintenance

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TABLE 2 '

CLASS 2 COMPONENTS NO RELIEF REQUESTS ,

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COMPONENT SUPPORTS LICENSEE PROPOSED

.F-2500-1 IWF-2500-1 SYSTEM OR ITEM TO BE REQUIRED ALTERNATIVE RELIEF REQUEST

'IM NO . EXAM. CAT. COMPONENT EXAMINED METHOD EXAMINATION STATUS i 30 F-A Reactor We ld & Visual None . Not Vessel Mechanical (VT-3) Granted Connections at Intermediate Joints in Multi-connected Inte-gral & Nonintegral Suppcrts (Reactor Vessel Support Skirt Weld) e e

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1 CONCLUSION 1 -

Based on the review summarized, we conclude that relief granted from the examination and testing requirements and alternate methods imposed through this document give reasonable assurance of the piping and component pressure boundary and support structural integrity, that granting relief where the

,/ Code requirements are impractical is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest considering the burden that could result if they were i

imposed on the facility. The relief request denied by the staff requires either compliance with Code requirements or submission of an alternative examination method to be evaluated by the staff to determine that adequate assurance of the structural integrity of the reactor vessel support skirt will be provided.

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NRC contributors to this SE: George Johnson, Guy S. Vissing Dated: January 13, 1986 l

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