ML20212J925

Safety Evaluation Accepting Licensee Relief Request IWE-3 for Second 10-year ISI for Plant

ML20212J925
Person / Time
Site:	North Anna
Issue date:	10/01/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20212J913	List: ML20212J910 ML20212J925
References
NUDOCS 9910060043
Download: ML20212J925 (3)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF NO. IWE-3 VIRGINIA ELECTRIC AND POWER COMPANY

' NORTH ANNA POWER STATION. UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339

1.0 INTRODUCTION

I'n the Federal Register dated August 8,1996 (61 FR 41303) the Nuclear Regulatory Commission (NRC) amended its regulations to incorporate by reference the 1992 Edition with

.1992 addenda of Subsections IWE and IWL of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. Subsections IWE and IWL provide the requirements for inservice inspection (ISI) of Class CC (concrete containment), and Class MC (metallic containment) of light-water cooled power plants. The j effective date for the amended rule was September 9,1996, and it requires the licensees to i incorporate the' new requirements into their !SI plans and to complete the first containment inspection by September 9,2001. However, a licensee may propose alternatives to or submit a request for relief from the requirements of the regulation pursuant to 10 CFR 50.55a(a)(3) and (g)(5). {

By letter dated August 4,1998 (Ref.1), the licensee, Virginia Electric & Power Company, submitted relief request Nos. RR-lWE-2, RR-lWE-3 (Unit 1 only), RR-lWE-4, RR-lWE-5, RR-lWE-6, and RR-lWL-2, seeking relief from some of the American Society of Mechanical

Engineers (ASME) Code,Section XI, Subsections IWE and IWL requirements for North Anna Power Station, Units 1 and 2. These relief requests have been submitted for the inspections to j be performed during the second 10-year inservice inspection (ISI) interval. By letter dated April 5,1999 (Ref. 2), the licensee withdrew RR-lWE-3. By letter dated April 21,1999 (Ref. 3), i the NRC staff granted the relief for RR-lWE-2, RR-lWE-4, RR-lWE-6 and RR-lWL-2; and authorized the use of the proposed alternative in RR-lWE-5. 1 In a letter dated September 16,1999, the licensee submitted Relief Requests IWE-3, IWE-7,  !

. and !WE-8. By letter dated September 27,1999 (Ref. 4), the licensee resubmitted the relief request RR-lWE-3. This evaluation addresses the acceptability of RR-lWE-3.

2.0 EVALUATION Reauest for Relief PR-lWE-3 Code Reauirement: Paragraph IWE-5240 of the 1992 Addenda of the Code requires that the requirements of Paragraph IWA-5240 for visual examinations (VT-2) are applicable following repair, replacement, or modification.

Enclosure 991 43 991001 POR K 05000338 P PDR l

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Licensee's Proposed Alternative: A pre-service visual examination (VT-3) will be performed on l the repair / replacement to determine the general mechanical and structural condition of the l repair. Testing and examination shall be conducted in accordance with 10 CFR 50, Appendix J, i as applicable. If a VT-2 visual examination cannot be performed in conjunction with the test,  ;

then a VT-1 visual examination will be performed to determine a more detailed condition of the repair after tne test is completed. i Staff Evaluation: IWE-5220 requires tosts following repair, modification, or replacements.

These tests include leakage tests in accordance with the provisicm of Title 10, Part 50 of the l Code of Federa/ Regulations, Appendix J, Paragraph IV.A, and visual examination VT-2 in  !

accordance with IWA-5040. The Appendix J tests allow for a certain amount of leakage.

However, in accordance with IWE-5250, if the Appendix J leakage rates are exceeded, the source of leakage chall be located and the area shall be examined to the axtent necessary to establish the requirements for corrective action. The licensee proposes, in lieu of a VT-2 visual examination, to perform a pre-service VT-3 visual examination on the repair or replacement. In )

- addition, the licensee proposes to perform a VT-1 visual examination after a test is completed, when a VT-2 visual examination is not feasible to perform. The use of a pre-service VT-3 visual examination.on the repair or replacement and a VT-1 when a VT-2 visual examination is not feasible provides reasonable verification that the structural integrity of the repair or replacement is sound and ensures the pressure boundary integrity of the containment. Therefore, the  ;

licensee's alternative proposal, to perform VT-1 visual examination following the test, where VT-2 visual examination in conjunction with the test is not feasible, is a reasonable approach for maintaining quality of a repair modification or replacement and for ensuring the integrity of the containment pressure boundary. I

3.0 CONCLUSION

Based on the information provided in the relief request, the staff concludes that the licensee's proposed alternative in RR-lWE-3 will provide an acceptable level of quality and safety.

Therefore, the proposed alternative is authorized for containment inspections pursuant to 10 CFR 50.55a(a)(3)(i).

Principal Contributo' r: H. Ashar Date: October 1, 1999 1 l

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j [Refe'rences :-

!- 1.- Letter from L. N. Hartz (VEPCO) to NRC, "ASME Section XI Relief Requests, North Anne Power Station, Units 1 and 2," August 4,1998

2. Letter from J. H. McCarthy.(VEPCO) to NRC, " Withdrawal of RR-lWE3 Relief Request,"

. April 5,1999 -

- 3. Letter from Richard Emch (NRC) to J. P. O'Hanlon (VEPCO), " North Anna Power Station, Unit 1 and Unit 2 - ASME Section XI Relief Requests, IWE2, IWE4 through IWE6 and IWL2 (TAC Nos. MA2476 and MA2477)," April 21,1999

4. Letter form L. N.' Hartz (VEPCO) to NRC," North Anna Power Stations Units 1 and 2, Surry Power Stations Units 1 and 2, ASME Section XI Relief Requests, Containment Liner IWE- -

3,7, and 8," September 16,1999 j

5. Letter from L. N. Hartz (VEPCO) to NRC, " North Anna and Surrey Power Stations Units 1 and 2: ASME Section XI Raiief Request tWE-3," September 27,1999.

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L J. P. O'Hanlon October 1, 1999 10 CFR 50.55a(a)(3)(i), the proposed alternative is authorized in that it provides an acceptable level of quality and safety. The staff's evaluation and conclusions are contained in the Enclosure.

Sincerely,.

ORIGINAL SIGNED BY:

Richard L. Emch, Jr., Chief, Section 1 Project Directorate il Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339 l

Enclosure:

As stated cc w/ encl: See next page Distribution l (Docket File-l PUBLIC l PDil R/F l H. Berkow E. Dunnington G. Edison R. Scholl (e-mail SE only)

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