ML20235S286
| ML20235S286 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/25/1987 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Wilgus W FLORIDA POWER CORP. |
| References | |
| NUDOCS 8710080413 | |
| Download: ML20235S286 (2) | |
Text
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s September 25, DB7 Docket No. 50-302 Mr. W. S. Wilgus Vice President, Nuclear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing P. O. Box 219 Crystal River, Florida 32629
Dear Mr. Wilgus:
SUBJECT:
ISSUES IDENTIFIED DURING THE OPERATIONAL SAFETY TEAM INSPECTION AT CRYSTAL RIVER UNIT 3 During our recent Operational Safety Team Inspection (OSTI) at Crystal River Unit 3, a number of strengths and weaknesses were identified, and discussed with you and your staff during the exit meeting on September 4, 1987. These will be presentea in our report of the inspection to be issued, and appropriate actions will be discussed at that time.
However, some of the issues are of such a nature that we believe they should be adequately resolved prior to restart from your present refueling outage.
We have discussed these matters with your staff, but the purpose of this letter is to document our concerns and to assure timely action.
It is our understanding that you are actively pursuing resolution of these issues.
Early and complete submittal of your proposed resolutions should preclude any delay in resumption of operations following the outage.
The first matter of concern relates to operation of the plant outside its design basis for maximum raw water (ultimate heat sink) temperature. The design basis document for the plant indicates that the maximum raw water temperature is 85 F.
This value was specified by Florida Power Corporation in the original design and was used as the basis for the accident analysis by the architect / engineer. During the inspection, the team observed actual raw water temperatures as high as 90 F.
Although a technical specification limiting condition for operation of 105 F exists for this parameter, the value of 105 F appears to have no basis.
A second issue concerns the assurance of adequate r.uclear services closed cycle cooling water (SW) system flows to safety-related motor coolers. During preoperational testirg, the system flow balance apparently established precise flow only to the reactor building fan coolers, spent fuel coolers, and the control complex chillers. Only indication of flow through other components was recorded. The SW system description indicates that the cooler outlet valves on the safety-related motors are occasionally throttled to acconnodate changes in water temperatures.
The tean was concerned that adequate component cooling may not be available during an accident if nominal flows have been established by the operators for conditions of normal plant operation.
Cooling for these motors appears to be the limiting factor for raw water 1
temperature.
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A third matter concerns the loading of the emergency diesel generators. There are several aspects to this concern. Technical Specification 4.8.1.1.2(D)(4) establishes an 18-month surveillance test requirement of greater than or equal to 3000 KW for greater than or equal to 60 minutes. The present 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of the diesel generator is 3000 KW and the 30 minute rating is 3300 KW, Therefore, the test conditions may exceed the rating of the machine.
Furthermore, the maximum automatically-connected load of the diesel generators is 3180 KW, but since 1977 diesel generator load testing was never greater than 3100 KW, which is less than the accident load requirement. Vendor correspondence reviewed during the inspection indicated that although the diesel was rated at 3000 KW for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> (upgraded to the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating reflected in the FSAR), this was for an ambient of 105'F. A 10% overload (3300 KW) was allowed for 30 minutes. The actual overload maximum rating for the Crystal
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River maximum ambient design temperature of 120 F was 3130 KW, below the maximum anticipated accident load.
Your staff has suggested a meeting in October to discuss these concerns and their resolution. We will make available appropriate staff members for that meeting to expedite our review of your submittals.
Sincerely,
/ 3/
Steven A. Varga, Director Division of Reactor Projects I/II cc: See next page Distribution:
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