ML18153A877

Requests That NRC Consider Rulemaking to Clarify Intent of 10CFR50.55a & Its Interface W/Asme Section XI as It Applies to Structural Integrity & Operational Leakage

ML18153A877
Person / Time
Site:	Surry, North Anna
Issue date:	02/28/1994
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
93-698, NUDOCS 9403090094
Download: ML18153A877 (5)

Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 28, 1994 United States Nuclear Regulatory Commission Serial No.93-698 Attention: Document Control Desk NL&P/ETS R3 Washington, D. C. 20555 Docket Nos. 50-280 50-281 50-338 50-339 License Nos. DPR-32 DPR-37 NPF-4 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 ASME SECTION XI and 10 CFR 50.55a INTERFACE The NRC has issued several documents, including Generic Letter (GL) 90-05 and Generic Letter 91-18, which provide guidance for implementation of ASME Code Section XI and 10 CFR 50.55a requirements. In addition to this generic guidance, the NRC has provided several interpretations of the Code and its implementation and interface with 10 CFR 50.55a on the subject of structural integrity and operational leakage in Code Class piping. These interpretations were provided to Virginia Electric and Power Company in your letter of October 22, 1993. In certain cases, we believe this guidance and interpretations impose a backfit of regulations without a commensurate cost benefit analysis. Also, it would appear that the ASME Code Committee and the NRC differ in their interpretation of the Code requirements for structural integrity/leakage referenced in GL 91-18 as evidenced by the attached Code interpretation.

Virginia Electric and Power Company's nuclear safety policy and operating philosophy are consistent with the intent of the GL 91-18. Our assessment of degraded or nonconforming conditions is based on their safety significance and Technical Specifications requirements. Specifically, for systems, structures, and components outside the scope of Technical Specifications, engineering assessments are used to evaluate safety significance and the need for corrective actions. During the evaluation process there must be a reasonable expectation that the systems, structures, or components are and will continue to remain operable (i.e., assurance that the capability exists for the system, structure or component to perform its specified function). If reasonable assurance cannot be demonstrated that the system can

(_9_4_0_3_0_9_0_0_9_4_*_94_0_2_2_8_ _ _ ;.~

. \

PDR. ADOCK 05000280 P PDR

perform its intended function, a JCO is developed or appropriate actions to place the plant in a safe condition are taken.

In contrast, the NRC tends to require additional actions that are largely administrative and only marginal to safety because it expects verbatim compliance with the guidance in these generic letters and its Code interpretations . For example, GL 90-05 requires the provisions of the generic letter to be met and NRC approval to be granted prior to making a temporary repair. The delays required to obtain prior-to-use relief requests from the ASME Code can be inconsistent with safe and prudent operation. In addition, we believe that the existing NRC guidance results in the unnecessary submittal of numerous relief requests for low energy Class 3 systems. These submittals present an undue burden on both NRC and utility resources without a significant increase in the safe operation of the affected system.

At the regional workshops held on GL 91-18 in 1992, questions were raised and concerns identified about numerous portions of the guidance. Based on this industry feedback and subsequent interpretations by the NRC on this subject of operational leakage in Code Class piping, we believe there was inadequate resolution and communication of the fundamentals of this issue. If the existing review process for generic communications had been in place before the issuance of GLs 90-05 and 91-18, we believe that the controversy and communication deficiencies would have been resolved prior to issuance. Based on the varying interpretations and implementation of the Code requirements for operational leakage, it is appropriate for the NRC to consider rulemaking to clarify the intent of 10 CFR 50.55a and its interface with ASME Section XI as it applies to structural integrity and operational leakage.

Very truly yours,

__-;c.j_

lit <C:(~~'1..-----------

) ~

W. L. Stewart Senior Vice President Nuclear Attachment

e cc: LJ;S; Nuclear Regulatory Commission Region II Suite 2900 101 Marietta Street, N.W.

Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Mr. R. D. McWhorter NRC Senior Resident Inspector North Anna Power Station Mr. W. T. Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. J. F. Colvin President and Chief Executive Officer Nuclear Management and Resource Council 1776 Eye Street N.W.

Suite 300 Washington, D. C. 20006-3706 Mr. W. H. Rasin Vice President and Director Technical Division Nuclear Management and Resource Council 1776 Eye Street N.W.

Suite 300 Washington, D. C. 20006-3706 Mr. T. E. Tipton Vice President and Director Operations, Management and Support Services Division Nuclear Management and Resource Council 1776 Eye Street N.W.

Suite 300 Washington, D. C. 20006-3706

A .

• e Attachment 1 ASME Code Inquiry

e ~ The American Society of

~ Mechanical Engineers

- Codes and Standards 345 East 47th Street New York. NY 10017 March 10, 1992 A. McNeill Virginia Power ISI/NDE Services 5000 Dominion Blvd Glen Allen, VA 23060

Subject:

Section XI, Table IWA-5210-1, IWA-5250 and IWB-3000, IWC-3000, IWD-3000; Corrective Measures - Prior to Continued Service (1980 Edition With the Winter 1980 Addenda, and Later Edition and Addenda Through Summer 1983 Addenda)

Item: IN92-005

Reference:

Your letter dated January 9, 1992

Dear Mr. McNeil!:

our understanding of the questions in your inquiry, and our replies are as follows:

Question (1): Does leakage identified during the conduct of a visual (VT-2) examination performed in conjunction with a Section XI required pressure test (Table IWA-5210-1), exceeding the acceptance criteria of IWB-3000, IWC-3000, and IWD-3000, require corrective measures in accordance with IWA-5250(a) prior to continued service?

Reply (1): Yes.

Question (2): Does leakage identified during the conduct of normal plant operation not in conjunction with a Section XI required pressure test (Table IWA-5210-1) require corrective measures in accordance with IWA-5250(a)?

Reply (2): No,Section XI, IWA-5250(a) does not apply during normal plant operation.

Very truly yours, Steve Weinman Assistant Secretary, Boiler & Pressure Vessel Committee ---

(212) 605-4720 ASME proceoures prov1oe tor recons1oeratron of this mterorelat1on when or 1f add1t1onal mtorma11on rs available which the inQmrer believes mront .1flec1 the mterpre1a11on Further. persons aqar1eved ti~* :r. 1~ ~1rrc:r~1a11on rnay :ioo~ai t:..1 1hr cognizant ASME comm11tee or subcomm,ttee. As stated 1n lhe foreword of Ille code documents. ASME does flat **approve ** *cert1I;*. ** rate. or **endorse

• any item construct,0:1 proor,erary oevrcP. or actrv1tv