ML18153A953
| ML18153A953 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 05/02/1994 |
| From: | Russell W Office of Nuclear Reactor Regulation |
| To: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| Shared Package | |
| ML18153A946 | List: |
| References | |
| GL-90-05, GL-90-5, NUDOCS 9405240136 | |
| Download: ML18153A953 (2) | |
Text
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e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 Docket Nos. 50-280, 50-281 50-33~ _an.d 50-339 Mr. W.L. Stewart Senior Vice President Nuclear Virginia Electric and Power Company Richmond, VA 23261
Dear Mr. Stewart:
MAY O 2 1111
SUBJECT:
SURRY AND NORTH ANNA POWER STATIONS - ASHE CODE INTERPRETATION AND RELIEF REQUEST PROCESS Thank you for your letter of February 28, 1994.
We have carefully reviewed your letter and we do not believe that your interpretation of our October 22, 1.993 1 etter is consistent with either the information contained in the 1 etter or its intent. With regard to moderate energy Class 3 systems, the process outlined in the letter provides a screening methodology for licensees to assess a degraded leaking pipe to determine if it can be considered operable.
The analytical methodology in Generic Letter (GL) 90-05, as well as another analytical method that was included in a proposed ASHE Code Case, in conjunction with the inspection guidance in GL 90-05, have been used by many licensees to determine if leaking piping in a moderate energy Class 3 system can be considered degraded but operable. If the guidance in the GL is satisfied, the licensee must submit a written relief request to the NRC within 30 days.
Any relief requested for leaking piping in ASME Code Class 1, 2, or high energy Class 3 systems must be approved by the NRC before it is implemented.
The above guidance is consistent with the staff position in GL 91-18.
While there may be certain instances where the required NRC action may not be consistent with cost-benefit consideration, the staff's experience with reviews of degraded piping has identified numerous instances where the degradation was safety significant. For example, in a recent review of leaking moderate energy Class 3 degraded piping, many additional portions of piping degraded by microbiologically induced corrosion were identified. The degradation in this system was so severe that a finding of reasonable assurance with regard to its structural integrity under design conditions could not be reached.
In the case of high energy systems, a licensee in the past has mischaracterized the mode of degradation associated with a leak, resulting in a situation where the potential existed for an unisolable blowdown of a steam generator.
In our previous letter to you, we stated that we were aware of Code Interpretation IN 92-005. Although we did not agree with the interpretation, we did not take exception to it as it only relates to the Code applicability, rather than a regulator.>' reguirement or position. However, we pointed out
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Mr. W. MAY O 2 199t that when a component is leaking it is necessary to assess its suitability for service. The flaw evaluation criteria in ASME Code,Section XI that are included in<<10CFRSO:,.-are not limited to assessing cracks in a component and do not permit through-wall defects.
Further, the Section XI flaw evaluation rules for assessing pipe-wall degradation do not permit through-wall degradation.
The evaluation methodology in GL 90-05 for moderate energy Class 3 piping and that which was contained in the proposed Code Case approved by Section XI was clearly a relaxation.
The prop~sed Code Case that had been used successfully to grant_
relief to many licensees did not receive the approval by the Main Cormnittee of the ASME Code because it was allegedly too conservative. If the proposed Code 1
Case had been approved and subsequently endorsed, the administrative burden of a submittal to the NRC for the degraded moderate energy Class 3 piping that satisfied the provisions of the Code Case would have been eliminated both for the licensee and the NRC.
Considering the above, the NRC staff agrees that rulemaking is appropriate to ensure plant safety and reduce unnecessary administrative burden, and has initiated such an activity.
We believe that a meeting with ASME Code personnel is warranted to discuss the proper relationship between ASME Code and NRC regulations.
We are exploring1 ways to set up a meeting between the ASME and the NRC.
Sincerely, Oric1nal Signed By-lILLUJI T. BUSSF.LD William T. Russell, Director Office of Nuclear Reactor Regulation Distribution WTRussell FJMiraglia LEngle NRC & Local PDRs FGillespie Central Files RHermann BBuckley HBerkow KBohrer-YT 0042
- SEE PREVIOUS CONCURRENCE
- DE:EMCB RAHermann:eh:adl:csp
- DE:EMCB JRStrosnider 03/25/94 04/14/94 i~,
D: NRalufJV' WTRussell
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