ML19291C447
| ML19291C447 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/10/1980 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8001240443 | |
| Download: ML19291C447 (2) | |
Text
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TOLEDO Docket No. 50-346 License No. NPF-3 R:cHAAD P. CaouSE Vice NS Ce"L Serial No. 575 N19F m n 39s221 January 10, 1980 Director of Nuclear Reactor Regulation Attention:
Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Reid:
Under separate cover, we are transmitting three (3) original and forty (40) conformed copies of an application for Amendment to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station Unit No.
1.
This application requests that the Davis-Besse Nuclear Power Station Unit 1 Technical Specification, Appendix A, be revised to reflect the change attached.
This proposed change allows a delay in the testing of the reactor vessel internal vent valves.
This amendment request involves a single change of the Class III type.
It is therefore determined to be a Class III amendment. Enclosed is $4,000.00 as required by 10CFR170.
The attachment identifies the proposed change, its safety evaluation and schedule required to implement the change af ter NRC approval.
Very truly yours, RPC:FRM: cts Attachments yn' i l/4
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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. CH!O 43652 8 0 01240 YY3 0
Docket No. 50-346 License No. NPF-3 Serial No. 575 January 10, 19 80 Change to Davis-Besse Nuclear Power Station Unit No. 1 Technical Specifications, Appendix A of Technical Specification 4.4.10.1 concerning reactor vessel internal vent valves.
See proposed change attached.
A.
Time Required to Implement -
This change can be ef fective upon NRC issurance, and i; required prior to March 27, 1980.
B.
Reason for Change (Facility Change Request 79-316)
See attached safety evaluation i794 175
Docket No. 50-346 License No. NPF-3 Serial No. 575 January 10, 1980 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
Each internals vent valve shall be demonstrated OPERABLE at least once per 18 months *during shutdown, by:
l 1.
Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation, 2.
Verifying the valve is not stuck in an open position, and 3.
Verifying through manual actuation that the valve is fully open when a force.of < 400 lbs. is applied vertical-ly upward.
- At the end of the first fuel cycle the testing and inspection may be delayed up to May 30, 1980.
j 1794 176 DAVIS-BESSE, UNIT 1 3/4 4-31 gooII803[]
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Docket No. 50-346 Page One of Two License No. NPF-3 Serial No. 575 January 10, 1980 SAFETY EVALUATION FOR DELAY IN TESTING OF REACTOR VESSEL INTERNALS VENT VALVES The biternals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internals vent valves fulfill the following purposes:
1.
Ensure valve operability
' 2.
Ensure that valves are not stuck during normal operation.
3.
Demonstrate that the valves are fully open at the forces equivalent to the differential pressure assumed in the safety analyses.
The inspection program for these valves also ensures that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, this program is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
Davis-Besse Unit 1 Technical Specification 4.4.10.1.b requires that operability of these valves be demonstrated at least once per 18 months during shutdown.
Technical Specification 4.0.2.a provides a maximum allowable extension of 4.5 months (25% of 18 months) for the 18 month period.
The valves were last tested' on May 12, 1978 during the BPRA/ ORA rt-Jal outage.
In order to meet the Technical Specification surveillance interval, the valve operability should, therefore, be demonstrated by March 27, 1980. This License Amendment Request provides for a relaxation in that schedule by approximately two months.
The testing is now proposed to be completed by May 30, 19 80.
The present scheduled date for the first refueling outage at Davis-Besse Unit 1 is March 15, 1980. The projected refueling schedule does not guarantee the fulfill-ment of Technical Specification surveillance requirement for the internals vent valves by March 27, 1980.
It should be noted that Davis-Besse Unit I has undergone several prolonged outages (including the March 30 to July 11, 1979 outage) af ter the testing done in May,1978.
This saving in core operating time is enough to account for the delay in testing s4.nce the intent of Section XI of ASME Boiler and Pressure Vessel Code is to provide a mechanism for the direction of an inspection based on operating _t_ime of the power facility.
Paragraph IWA-2400(a) of the 1974 Section XI with addenda through Summer 1975 states:
"...These inspection intervals represent calendar years af ter the reactor facility has been placed into commercial service.
The interval may be extended by as much as one year to permit inspections to be con-current with plant outages."
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Docket No. 50-346 License No. NPF-3 Page Two of Two Serial No. 575 January 10, 19 80 In addition to the above, B&W has informed Toledo Edison (letter DB-79-223 dated November 20, 1979) that an extension in the required technical specifi-cation for exercising the valves by several months is justified. This is based on successful demonstration of valve exercise at ANO-1, SMUD, Crystal River-III and Oconee-l.
Pursuant to the above, if the scheduled date for completion of valve testing is extended to May 30, 1980:
- 1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
- 2) A possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.
- 3) The margin of safety as defined in the basis of any Technical Specification is not reduced.
Consequently, this is not an unreviewed safety question.
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