ML20137L521
| ML20137L521 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/22/1986 |
| From: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8601280009 | |
| Download: ML20137L521 (1) | |
See also: IR 05000312/1985030
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Ji\\N 2 21999
Docket No. 50-312
Sacramento Municipal Utility District
P. O. Box 15830
Sacramento, California 95813
Attention:
Mr. R. J. Rodriguez
Executive Director for Nuclear Operations
Gentlemen:
Thank you for your letter dated December 23, 1985, informing us of the steps
you have taken to correct the items which we brought to your attention in our.
letter dated November 20, 1985. Your corrective actions will be verified
during a future in;pection.
Your cooperation with us is appreciated.
Sincerely,
.
hek
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Dennis F. Kirsch, Acting Director
Division of Reactor Safety and Projects
bcc w/ copy of letter dated 12/23/85:
J. Martin
B. Faulkenberry
G. Cook
RSB/ Document Control Desk (RIDS)
Resident Inspector
Project Inspector
State of CA
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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street. P.O. Box 15830. Sacramento CA 95852-1830.(916) 452 3211
AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA
RJR 85-595
December 23,1985
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REGION V 0FFICE OF INSPECTION AND ENFORCEMENT
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NOTICE OF VIOLATION FOR NRC INSPECTION 85-30
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The Sacramento Municipal Utility District hereby submits in Attachment I to
this letter, a response to the subject Notice of Violation in accordance with
If there are any questions concerning this response, please
act Mr. Ron W. Colombo at the Rancho Seco Nuclear Generating Station.
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R. J.
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ASSISTANT GEN AL MANAGER,
NUCLEAR
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ATTACHMENT I
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DISTRICT RESPONSE TO NRC INSPECTION 85-30
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-
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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
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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 2.51
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.
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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 IX. 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.
B.
Technical Specification 3.1.2.2 states, in part "...Heatup and cooldown
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rates chali G t exceed the rates stated on the associated figure." Figure
3.1.2- ) states a reactor coolant system cooldown rate of 100 F/Hr at
temperatures greater than 270F.
~ On October 2, 1985, with the reactor at approximately 15 percent power, a
loss of main feedwater caused a high reactor coolant system pressure,
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resulting in a reactor trip. Primarily due to the inappropriate opening
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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 20 minutes.
This cooldown exceeded the Technical Specification limit as specified
above and is a Severity Level IV Violation (Supplement I).
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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/48 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.
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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 cooldewn 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
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should be further reduced to compensate for this condition. Following this
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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 steam '#ill be
supplied to the 4th point feedwater heaters without lifting the relief valves
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following a plant trip.
Full compliance was achieved with the' successful completion of this test on
December 7, 1985.
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