ML20077N817
| ML20077N817 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/12/1983 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Starostecki R, Starostecki R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20077N805 | List: |
| References | |
| NUDOCS 8309130227 | |
| Download: ML20077N817 (4) | |
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BALTIM ORE GAS AND ELECTRIC CHARLES CENTER P. O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR.
VicE PRESIDENT August 12, 1983 U. S. Nuclear Regulatory Commission Docket Nos. 50-317 Region I 50-318 631 Park Avenue King of Prussia, PA 19406 License Nos. DPR-53 DPR-69 ATTN:
Mr. Richard W. Starostecki, Director Division of Project and Resident Programs Gentlemen:
This refers to Inspection Report 50-317/83-13; 50-318/83-13, which transmitted several items of apparent non-compliance with NRC requirements.
Enclosure (1) to this letter is a written statement in reply to those items noted in your letter of July 14, 1983.
Should you have any further questions regarding this reply, we will be pleased to discuss them with you.
Very truly yours, f/
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J. A. Biddison, Esquire G. F. Trowbridge, Esquire Mr. D. H. Jaffe, NRC Mr. R. E. Architzel, NRC 8309130227 830907 PDR ADOCK 05000317 G
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ENCLOSURE (1)
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-REPLY TO APPENDIX A OF NRC INSPECTION h
REPORT 50-317/83-13; 50-318/83-13 y
ITEM A.1 As a result of the violation associated with operation of the Fire Pump against a shutoff head, Surveillance Test Procedure (STP) M-76-0, Staggered Test of the Electric Fire Pump, has been revised to incorporate the performance of surveillance testing using a recirculation flow path provided by n cans other than relief valve prctection. The fire pump in question is protected by a circulation relief valve and prior to the finding, STP M-76-0 was performed by running the pump with the discharge valve shut and allowing the circulation relief valve to cycle (with a setpoint below the shutoff head) as required to protect the pump. This method of testing provides an acceptable protective flowpath, is recognized by NFPA Code 20 (Section 2.6), and we consider this method to be the preferred method for testing the pump. As a result of a certain ambiguity associated with interpretation of the current Technical Specification surveillance requirements, we plan to submit an amendment to clarify the surveillance requirements and allow performance of the test per the previously described methods.
For the interim, the recent revision to STP M-76-0 brings us into full compliance with the Resident inspector's interpretation of the surveillance requirements contained in the current Technical Specifications. To ensure that similar violations do not occur in the future, while providing maximum operating flexibility, the following actions will be takem (1) A change to the Technical Specifications will be initiated to clarify any ambiguity regarding testing methodology, and (2) The circulation relief valve will be placed on a preventive maintenance schedule to ensure acceptable operation if used in future surveillance testing.
ITEM A.2 In response to the violation cited, the closed Component Cooling Water (CCW) isolation valve to #12 Hi;h Pressure Safety injection (riPSI) Pump Seal Water Cooler was opened.
Additionally, all other component cooling water isolation valves associated with the Unit I and Unit 2 HPSI Pumps were verified open immediately upon discovery of the closed valve for #12 HPSI Pump. The Inspection Report expressed concerns that inadequate controls currently exist to prevent the mispositioning of the cooling water supply valve to #12 HPSI Pump Seal Water Cooler and similar valves associated with the other redundant pumps.
The Inspection Report noted that the Unit I and 2 Operating Instructions (Ol's) for the CCW system failed to list the subject valve in the valve checklists associated with the OI. Contrary to the finding presented in the narrative section of the Resident inspector's Report, the Unit 2 CCW 01 (01-16) dated January 19, 1983, did include the CCW isolation valves to the Unit 2 HPSI Pumps (these valves were verified open). The Unit 1 CCW OI, however, did not include the corresponding CCW isolation valve. This 01 has subsequently been corrected.
We believe that the failure to control the Unit i valve in the proper position was as a result of an inadequate description in the OI valve checklist.
Our administrative mechanism for controlling these valves consists of routine maintenance tagging procedures and refueling cycle valve line-ups. The 01 checklists provide a means of identifying proper valve locations and positions whenever system realignment must be performed for maintenance.
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To ensure no other problems such as that associated with #12 HPSI go unidentified, we are performing valve line-ups using valve lists developed during recent system walkdowns. The walkdowns that have been performed represent a substantial manpower expenditure aimed towards reverifying plant system prints with the as-built configuration of the plant. We anticipate that the physical valve line-ups utilizing revised OI valve lists will be completed for Unit 2 accessible valves by September 30,1983, and for Unit i valves prior to startup from the October 1983, refueling outage. Line-up of the Unit 2 inaccessible valves will be completed during a subsequent outage of sufficient duration, but no later than start-up following the next scheduled Unit 2 refueling outage.
These actions provide assurance that Calvert Cliffs will continue to operate in compliance with Technical Specification 6.8.1.a and also provide the necessary corrective actions to ensure that violations of similar nature will not be repeated in the future.
ITEM A.3 In response to the 10 CFR 50.59 violation cited, the following action will be taken; a revised safety analysis will be prepared to evaluate and justify previous safety analysis performed to remove safety features signals from the sample isolation valves in question and redesignate valve functions for the post accident sampling system. This action will be completed by no later than November 1,1983.
Subsequent to the review and approval of previous safety analysis that evaluated changes to the sample valves in question and in response to the NRC's Performance Appraisal Inspection of January 1982, we have undertaken an extensive and continuing training program to review with all responsible engineers the depth and detail that safety analysis must include. Implementing procedures have been revised to include detailed discussion of how the safety analysis are researched, analyzed, and written. This has resulted in significant improvement in the socpe and thoroughness of current safety analysis and has been noted by NRC Inspectors.
Certain programmatic changes have been completed in response to the 10 CFR 50.59 violation. Among the more significant changes, the design control procedures have been revised to be more specific with regard to identification of Facility Change Requests (FCR's) that could affect the Technical Specifications. Greater emphasis has been placed on the safety analysis implementing procedures to ensure that any changes to the facility that affect either the margin of safety as defined in the Technical Specification bases or result in a change to any individual Technical Specification are properly identified. To prevent recurrence of violations of a similar nature and clarify 10 CFR 50.59 requirements, the Supervising Engineer-Nuclear Generation has reviewed with all EED responsible engineers the requirements necessary to ensure that no modified system involving changes to *he Technical Specifications or accompanying bases is declared operable and returned to service prior to the NRC approval of such changes. This is accomplished through the controls mentioned below.
The Responsible Design Organization's implementing procedures for control of changes, tests, and experiments has been revised to include a specific checkpoint to document review of the Technical Specifications.
A clarification has been added to require a separate Technical j
Specification Amendment FCR if a change to the Technical Specifications is needed relative to a facility change.
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3 In addition to the above action, procedures have bcen revised that govern on-site review and approval of all FCR's. This revision includes documenting an independent on-site engineering review of physical change FCR's for Technical Specification applicability.
These actions, combined with the greater awareness that has been placed on training involving preparation of safety analysis and review of hardware changes for Technical i
Specification applicability should be sufficient to preclude recurrence of similar violations in the future.
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