ML18153C914

Responds to NRC 920203 Ltr Re Violations Noted in Insp Repts 50-280/91-37 & 50-281/91-37,denying Violation Occurred.Prior to Restoring Leaking Letdown Line to Service,Licensee Found That Leakage Acceptable Per TS 3.1.C.3 Limits

ML18153C914
Person / Time
Site:	Surry
Issue date:	02/28/1992
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
92-104, NUDOCS 9203050213
Download: ML18153C914 (5)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 28, 1992 U.S. Nuclear Regulatory Commission Serial No.92-104 Attn: Document Control Desk SPS/RJS/ETS R5 Washington, D.C. 20555 Docket Nos.: 50-280 License Nos.: DPR-32 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 1 (

REPLY TO A NOTICE OF VIOLATION NRC INSPECTION REPORT NOS. 50-280/91-37 AND 50-281/91-37 We have reviewed your Inspection Report Nos. 50-280/91-37 and 50-281/91-37 dated February 3, 1992 and the enclosed Notice of Violation. The violation identified a failure to submit requests for evaluation and relief from the requirements of IWA-5250 of ASME

• Section XI in accordance with 10 CFR 50.55a(g)(6)(i). We deny that a violation occurred based upon our interpretation of the ASME Section XI Code, the Surry Technical Specifications, and the Standard Technical Specifications for Westinghouse PWRs, NUREG-0452 Revision 4.

The cover letter of this inspection report stated that "Recent NRC guidance in this area, which was communicated to you in Generic Letter 91-18, clearly indicated that relief from the code requirements is required in this type situation." We do not agree. ASME Section XI Code Committee Inquiry IN91-017, dated September 10, 1991, stipulated that subparagraph IWA-5250 does not apply to operational leakage. This position was reconfirmed during an ASME Section XI Code Committee meeting on February 12, 1992. In addition; *the Standard Technical Specification surveillance requirements and actions in the structural integrity section are consistent with this interpretation.

The Surry Technical Specifications provide the requirements for operational leakage.

Surry Technical Specification 3.1.C.3 limits identified and isolable leakage to 1O GPM.

System operability was considered prior to restoring the leaking letdown line to service. Subsequently, a safety evaluation was performed to document the issues considered during the course of the evaluation, and it concluded that continued operation with the letdown leakage was acceptable.

Because it was clearly understood that a permanent code repair to the weld would have to be made, or relief requested in accordance with GL 90-05, the letdown line was returned to service as a prudent operational measure while the code repair plan was being finalized. We do not believe that restoration of the letdown line to service in and of itself constitutes a "temporary repair," nor do we believe that Generic Letter 90-

• 05 intended such a restoration to be considered a temporary repair as evidenced by the examples of temporary repairs discussed in the Generic Letter.

9203050213 920228 PDR ADOCK 05000280 G PDR

In summary, we believe we executed a timely code repair to minor identified and isolable operational leakage that was within Technical Specification limits. We also evaluated the safety consequences of continued operation with such leakage and kept the NRC fully informed of our decisions throughout the repair process. In addition, we responded to several design specific questions raised by the NRC resident inspectors during the course of the repairs. A more detailed discussion of the sequence of events surrounding the repair is included in Attachment 1.

Based upon the well established ASME position on operational leakage, we are concerned that the GL 91-18 position on operational leakage represents a backfit which has not been completely evaluated. Furthermore, should such a change in position on operational leakage be considered necessary by the NRC, we believe that its consistent implementation warrants rule making vice dissemination of guidance in an NRG Inspection Manual. We believe that implementation of this new position is not commensurate with any regulatory benefit and could have an adverse impact on the safe and reliable operation of nuclear power plants, by creating the potential for unnecessary transients or evolutions to repair non-safety significant leakage.

Should you have any questions concerning our response, please contact us. We have no objection to this reply being disclosed to the public.

Very truly yours,

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W. L. Stewart Senior Vice President Nuclear Attachment cc: U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station

REPLY TO A NOTICE OF VIOLATION NRC INSPECTION CONDUCTED DECEMBER 1, 1991 - JANUARY 4, 1992 SURRY POWER STATION UNIT 1 INSPECTION REPORT NOS. 50-280/91-37 *AND 50-281/91-37 NRC COMMENT:

During the Nuclear Regulatory Commission (NRC) inspection conducted on December 1, 1991, through January 4, 1992, a violation of NRC requirements was identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1991 ), the violation is listed below:

10 CFR 50.55a(g)(6)(i) states in part that the Commission will evaluate determinations of Code impracticality and may grant,. relief and may impose alternate requirements.

Contrary to the above, on December 20, 1991, the licensee returned the Unit 1 letdown system to service without repairing a leak on a weld of the Class 2 letdown line. This leak did not meet the Code requirements in IWA-5250 of Section XI of the ASME Code and the licensee did not submit requests for evaluation and relief from the Code requirements.

This is a Severity Level IV violation (Supplement I) .

REPLY TO A NOTICE OF VIOLATION NRC INSPECTION CONDUCTED DECEMBER 1, 1991 - JANUARY 4, 1992 SURRY POWER STATION UNIT 1 INSPECTION REPORT NOS. 50-280/91-37 AND 50-281/91-37 (1) Reason for the Violation, or. if Contested. the Basis for Disputing

. the Violation On December 13, 1991, with Unit 1 at 100% reactor power, an increase in unidentified Reactor Coolant System (RCS) leakage from 0.28 gpm to 0.37 gpm was recorded during the performance of PT-10, "Reactor Coolant Leakage."

Although this value was well below the Technical Specification limit of 1.0 gpm unidentified leakage, the frequency of the leakage monitoring was increased to several times per day. On December 16, *1991, two containment entries were made to locate the source of the leakage. The leakage was determined to be the result of a socket weld leak in the letdown line downstream of the letdown isolation valves. A station deviation report was initiated.

In order to repair the weld, extensive preparations were required for personnel safety due to the Surry subatmospheric containment and the local radiation levels in the area of the leak. On December 19, 1991 at 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br />, following preparations to isolate, drain, and repair the weld, Excess Letdown was placed in service and an initial attempt to isolate the leaking weld was made. Due to seat leakage through the intended isolation valves, this initial attempt to isolate the leak was unsuccessful. Because the next available isolation valve was located in the "A" loop room in a 2 rem/hour radiation field and had been back-seated for an extended period of time, it was determined that additional time would be required to evaluate and develop an alternate repair isolation plan.

Isolation of Normal Letdown removes both hydrogen addition capability and cleanup demineralizers for the RCS from service. Since preparation of the second repair plan was anticipated to require additional time and RCS hydrogen concentration was trending toward the low end of its band, it was desirable to restore Normal Letdown to service. Technical Specification system operability requirements and applicable code requirements were reviewed with Engineering. It was concluded that Normal Letdown could be returned to service with the existing operational leakage, and remain in service while the new permanent repair plan was being prepared. Based upon this evaluation, Normal Letdown was returned to service at 1517 hours0.0176 days <br />0.421 hours <br />0.00251 weeks <br />5.772185e-4 months <br /> on December 20, 1991.

Subsequently, a safety evaluation was performed to document the issues considered during the course of the above discussions, which concluded that an unreviewed safety question did not exist.

During the course of this review, the need for requesting relief from ASME Section XI requirements was considered, and it was determined that the code was not applicable to this operational leakage. The section of code which was cited in the Notice of Violation was IWA-5250. IWA-5250(a) describes corrective action to be taken for leakage discovered during the performance of

ASME Section XI pressure testing. We do not consider this applicable to leakage discovered during plant operations. Our interpretation has previously been confirmed by the ASME Section XI Code Committee Inquiry IN91-017 dated September 10, 1991, and was reconfirmed during a recent ASME Section XI Code Committee meeting on February 12, 1992. Section IWA-5250(b) is not applicable because the piping system is not made of ferritic steel.

Thus, we believe that this event does not constitute a violation of NRC requirements.

On December 24, 1991 at 0824 hours0.00954 days <br />0.229 hours <br />0.00136 weeks <br />3.13532e-4 months <br />, following completion of the repair plan, Excess Letdown was again placed in service and Normal Letdown isolated.

The leaking weld was permanently repaired in accordance with the*

requirements of the ASME Code on December 26, 1991 at 1922, and Normal Letdown was returned to service at 0053 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> on December 27, 1991.

(2) Corrective Steps Which Have Been Taken and the Results Achieved No violation occurred, and therefore no corrective steps were taken.

(3) Corrective Steps That Will be Taken to Avoid Further Violations No violation occurred, and therefore no corrective steps are planned.

( 4) The Date When Full Compliance Will be Achieved Virginia Power considers that its actions were in full compliance with 10 CFR 50.55a(g)(6)(i) and the Surry Technical Specifications.