ML20137H936
| ML20137H936 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/23/1985 |
| From: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| RJR-85-595, TAC-59691, NUDOCS 8601220147 | |
| Download: ML20137H936 (3) | |
Text
b' 6
,g f0-7 d
.esuu.
SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, P.O. Box 15830, Sacramento CA 958521830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA RJR 85-595 December 23,1985 r,
0s b
E, J B MARTIN REGIONAL ADMINISTRATOR g
REGION V 0FFICE OF INSPECTION AND ENFORCEMENT
,9 N
PM d $
U S NUCLEAR REGULATORY COMMISSION O
1450 MARIA LANE SUITE 210 5
n c
WALNUT CREEK CA 94596 1~.
r 0
DOCKET NO. 50-312 LICENSE N0. DPR-54 NOTICE OF VIOLATION FOR NRC INSPECTION 85-30 The Sacramento Municipal Utility District hereby submits in Attachment 1 to this letter, a response to the subject Notice of Violation in accordance with 10 CF 2.201.
If there are any questions concerning this response, please a
Mr. Ron W. Colombo at the Rancho Seco Nuclear Generating Station, ty R. J.
DRIG ASSISTANT GENT AL MANAGER, NUCLEAR Attachments
$f $
k
) \\
O rt-a/
^
ATTACMENT I DISTRICT RESPONSE TO NRC INSPECTION 85-30 NOTICE OF VIOLATION As a result of the inspection conducted by Mr. J. Eckhardt and Mr. G. Perez between September 28 and October 31, 1985, the following two (2) violations were identified.
Each violation is followed by the District's response to the violation.
A.
Technical Specification 6.8.1 requires that written procedures be es-tablished, implemented, and maintained covering activities including surveillance and test activities of safety related equipment. Also, Administrative Procedure 2 " Review, Approval and Maintenance of Pro-cedures" establishes the licensee's guidelines for revisions or temporary changes to procedures.
Contrary to the above requirements, on September 29, 1985, the licensee used an unapproved procedure in performing the calibration of the A channel power range instrument.
DISTRICT RESPONSE:
On September 29, 1985, Instrumentation and Control (I&C) Procedure I.103,
" Power Range Calibration," was being performed with the reactor power level at approximately 2.5%
The power level was not sufficient to allow the pro-cedure to be performed as written.
The I&C technician, using outside instruc-tions in violation of the procedure, changed a multiplier link From 10X tn IX so that Section 6.3 of I.103 could be completed.
The event was observed by a NRC resident inspector and reported to the shift supervisor.
The I&C technician, upon realizing his violation of the procedure, prepared an internal Occurrence Description Report (AP.22) to alert plant management personnel of the concern.
The immediate action taken to address the procedure violation was to incorporate a temporary change in Revision 7 of I.103 to include instructions describing how to change the multiplier link from 10X to 1X.
This action was completed on September 29, 1985. A permanent revision to I.103 is being prepared to in-clude the multiplier link change instructions.
The procedure revision will be completed and full compliance achieved by February 28, 1986.
-~
Technical Specification 3.1.2.2 states, in part "...Heatup and cooldown B.
rates shall not exceed the rates stated on the associated figure." Figure 3.1.2-2 states a reactor coolant system cooldown rate of 100 F/Hr at temperatures greater than 270*F.
On October 2, 1985, with the reactor at approximately 15 percent power, a loss of main feedsater caused a high reactor coolant system pressure, resulting in a reactor trip.
Primarily due to the inappropriate opening of the fourth point feedwater heater relief valves, the reactor coolant temperature decreased to below the normal post reactor trip temperature of approximately 550 F to approximately 490 F in 2J minutes.
This cooldown exceeded the Technical Specification limit as specified above and is a Severity Level IV Violation (Supplement I).
m J B MARTIN DISTRICT RESPONSE:
The District has evaluated the October 2, 1985 reactor trip and agrees that ex-ceeding the Technical Specification 3.1.2-2 reactor coolant system cooldown rate was primarily caused by the inappropriate lifting of the 4A/4B heater steam relief valves. When the unit is tripped, pegging steam is supplied to the 4A/4B heaters to provide preheating to the feedwater. The pegging steam pressure controller should control the pressure at a point lower than the relief valve setpoints.
It was found that the controller setpoints were adjusted so that the controller dead band actually overlapped the relief valve setpoints.
This led to the lifting of the relief valve and subsequent cool-down.
To address this concern, the relief valve setpoints were raised and the pegging steam pressure controller setpoints were reduced accordingly.
A sub-sequent reactor trip on December 5, 1985, resulted in lifting the relief valves and revealed that the concern had been only partially resolved, even though the cooldown rate was consistent with normal post-trip response. Although the setpoints were found to be satisfactory, it was found that the controller had an overshoot characteristic that had not been taken into consideration.
An engineering evaluation concluded that the pegging steam controller setpoint should be further reduced to compensate for this condition.
Following this final setpoint adjustment, functional test STP.187, "2nd, 3rd, 4th point FWH Relief Valve Test," was performed.
STP.187 tests the relationship between the pegging steam controllers and the opening of the 4A/4B heater steam relief valves.
The test demonstrated that, with the final setpoints, pegging staam will ba supplied to the 4th point feedwater heaters without lifting the relief valves following a plant trip.
Full cciapliance was achieved with the successful completion of this test on December 7, 1985.
i i
\\
I 1
i r