ML20069H248
| ML20069H248 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/15/1982 |
| From: | Larson C NORTHERN STATES POWER CO. |
| To: | Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20069H233 | List: |
| References | |
| NUDOCS 8210190439 | |
| Download: ML20069H248 (2) | |
Text
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/9 Northem States Power Company 414 Nicollet Mall Minneapoks. Minnesota 55401 Telephone (612) 330-5500 September 15, 1982 D
Mr. J.F.
Streeter, Chief Projects Branch 2 Region III Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Streetsr:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS. 50-282 AND 50-306 In response to your letter of August 17, 1982, which transmitted Inspection Reports 50-282/82-11 and 50-306/82-11, the following is offered:
Violation 10 CFR 50.72(a) states, in part:
"Each licensee of a 1
nuclear power reactor licensed under Sections 50.21 or 50.22 shall notify the NRC Operations Center as soon as possible and in all cases within one hour by telephone of the occurrence of any of the following significant events and shall identify that event as being reported pursuant to this section.
(7) Any event resulting in manual or automatic actuation of Engineered Safety Features, including the Reactor Protection System."
Contrary to the above, an inadvertent Engineered Safety Features actuation that occurred during the performance I
of Surveillance Procedure SP-2547 on July 7, 1982, was not reported to the NRC Operations Center.
This is a Severity Level V violation (Supplement I).
Response
The inadvertent actuation of one train of Safety Injection
}
occurred during routine testing while the unit was in
(
refueling shutdown.
Many components normally actuated by B210190439 821007
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l PDR ADDCK 05000282 i
e Mr. J. F. Streeter September 15, 1982 the Safety Injection were out of service because they need not be operable during refueling.
Due to plant conditions at the time of the actuation, no thermal transient was possible on the RCS or support systems.
1k) overpressure transient would result, and no loss of decay heat removal capabilities would occur.
No addition of positive reactivity could occur, and there would be no challenge to containment integrity.
It is felt that partial actuation of the engineered safeguards under these plant conditions does not constitute a significant event, does not meet the intent of 10 CFR 50.72 (a), and need not be reported.
Violation Technical Specification 3.10.B.5 states:
"Above 90 percent of rated thermal power:
If the indicated axial flux difference of two operable excore channels deviates from its target band, within 15 minutes either eliminate such deviation, or reduce thermal pcwer to less than 90 percent of rated thermal power."
Contrary to the above, on July 2, 1982, thermal power was not reduced to less than 90 percent of rated thermal power within 15 minutes after the indicated axial flux difference of two operable excore channels deviated from the target band.
The condition existed for about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before the
, condition was recognized and corrected by the licensee.
This is a Severity Level V violation (Supplement I).
Response
Upon discovery, flux difference on a)1 four channels was returned to the target band, power was reduced to less than 50%, and the power range high flux trip was reduced to less than 55%.
Changes to the startup program are contemplated which will allow normalization of the flux difference channels at part power based on incore flux mapping.
These changes will be made before Unit 1 is restarted from its fall 1982 refueling outage.
U C. E. Larson Director, Nuclear Generation CEL/ts cc Messrs. G. Charnoff C.D. Feierabend
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