05000482/LER-1985-046, Informs That Change in Commitment Associated to LER 85-046-00 to Develop Program to Periodically Check SG Feedwater Check Valve Leakage Rates

Informs That Change in Commitment Associated to LER 85-046-00 to Develop Program to Periodically Check SG Feedwater Check Valve Leakage Rates

ML20078B860
Person / Time
Site:	Wolf Creek
Issue date:	10/20/1994
From:	Rhodes F WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-94-0072, ET-94-72, NUDOCS 9410270122
Download: ML20078B860 (3)

LER-2085-046, Informs That Change in Commitment Associated to LER 85-046-00 to Develop Program to Periodically Check SG Feedwater Check Valve Leakage Rates

Event date:
Report date:
4822085046R00 - NRC Website

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• I W8LF CREEK ' NUCLEAR OPERATING CORPORATION Forrest T. Rhodes Vice President Engineering ,'

October 20, 1994 ET 94-0072 -

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station F1-137 )

Washington, D. C. 20555 l

Reference:

Letter KMLNRC 85-185 dated July 24, 1985 from G. L. Koester, KG&E, to NRC i

Subject:

Docket'No. 50-482: Change in Commitment Associated to Licensee Event Report 85-046-00 Gentlemen:

This letter provides a change in commitment associated to Licensee Event  !

Report (LER) 85-046-00. In response to an Engineered Safety Feature actuation event which was reported .in LER 85-046-00, a commitment was made to develop a program to periodically check the Steam Generator Feedwater Check Valve leakage rates. This commitment has been reevaluated and WCNOC has determined that other maintenance and inspection programs will provide assurance that the valves will fulfill their design function. The attachment ' to this letter provides further discussion and evaluation of this change in commitment.

If you have any questions concerning this matter, please contact me at (316)364-8831 extension 4002 or Mr. Richard D. Flannigan at extension 4500.

Very truly yours, 0 /Y r / -

Forrest T. Rhodes FTR/j ra

((] } Attachment cc: L. J. Callan (NRC), w/a D. D. Chamberlain (NRC), w/a J. F Ringwald (NRC), w/a J. C. Stone (NRC), w/a I PO Dox 411/ Burlington, KS 66839 / Phone: (316) 3644831 r 9410270122 941020 PDR An Equal opportunity Ernployer M/F/HC/ VET - g P ADOCK 05000482 PDR a

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Attachment'to ET- 94-0072 Page le of 2 Evaluation of Change to Commitment Associated to the Surveillance - of Steam Generator Feedwater Check Valve Leakage LER 85-046-00 was generated to report the occurrence of a steam generator drain down to the condenser by way of the feedwater system. The evaluation of this event indicated the root cause of the event to be the feedwater system lineup. During this event the Feedwater Isolation Valves (FWIVs) were maintained open while steam generator pressure was slightly above the feedwater header pressure. This operating condition allowed flow from the steam generator to the condenser by way of the feedwater system (reverse flow through the !: team Generator Feedwater Check Valve). ' Subsequent to this transient the operating procedures were revised to prevent re-occurrence of this condition by ensuring the FWIVs are closed when steam generator pressure is above the feedwater system pressure. In addition, WCNOC committed to test the Feedwater Check Valves to ensure operability by measuring back-leakage. This change will delete the commitment made in LER 85-046-00 to test the subject Feedwater Check Valves for back-leakage on a periodic basis. Results of the periodic testing have shown no negative trends in the .Ives' performance. These check valves are currently part of WCNOC's Check Valve Reliability Improvement program. This program as well as the ASME Inservice Testing program, ensures internal inspections, performance monitoring and non-intrusive testing are performed on Feedwater Check Valves. Internal ' inspections, open/ closure testing and non-intrusive testing of the Feedwater Check Valves provide greater operability and reliability information. These program requirements also. provide greater assurance that the Feedwater Check Valves will perform their intended design function than would be detected by reverse flow testing. An evaluation has been performed to confirm the acceptability of removing the requirement to test Feedwater Check Valves V120, V121, V122, and V123 for back leakage. During the event the leakage rate through the check valve was  ! determined to be well below the values specified in the loss of feedwater  ; accident analysis. However, this evaluation assumes that all four Feedwater Check Valves fail to seat, allowing back leakage and steam generator drain down to the FWIVs. This justification is based on evaluations involving the Updated Safety Analysis Report (USAR) Chapter 15 analyses requiring auxiliary

• feedwater to provide long-term core cooling (i.e., Feedwater Line Break, Loss of Normal Feedwater, and Loss Of Non-Emergency AC). The evaluation shows that current analyses for these events bound the cases in which excessive steam generator mass depletion due to back leakage through the main feedwater line check valves is considered. The evaluation also confirms that there is no increase in transient specific calculations of fuel rod failure. The evaluation and analyses support the conclusion that all safety analysis acceptance criteria continue to be met.
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f4 Attachment'to ET 94-0072 Page 2*of 2 Since this evaluation has been shown to produce less limiting results than the current USAR analyses, the probability of occurrence and the consequences of an accident evaluated previously in the USAR are . not increased and the proposed test change does not affect any of the mechanisms postulated'zn the USAR-to cause LOCA or non-LOCA design basis events. The evaluation confirms that the USAR conclusions remain valid for the proposed change since the current analyses have been shown to remain bounding. On these bases it is concluded that the probability and consequences of the accidents previously evaluated in the USAR are not increased. Therefore, it is-also concluded that tnere is no reduction to the currently analyzed margin of safety. 1 l l 1 1 l i 1 l l I

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