ML19309B024
| ML19309B024 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/08/1975 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19309B021 | List: |
| References | |
| NUDOCS 8004020501 | |
| Download: ML19309B024 (8) | |
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1.2 REPORT FOR fEBRlM Y
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1.2.1 OPERATIONS
SUMMARY
1.2.1.1 Changes in Design There were no significant changes in design of Rancho Seco Unit I made during February, 1975.
1.2.1.2 Performance Characteristics Primary ef fort this month was maintenance of the 92.5% power plateau and first attempt at the 100 hr. run. The effort was terminated on February 27th with occurrence of a condenser tube leak. The loss of off-site power test was also completed.
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1.2.1.3 Changes in Operating Methods The following significant safety related changes were made to Rancho Seco Unit 1 Operating Procedure in February, 1975:
1.
Procedure A.25, " Nuclear Service Raw Water System" Revision 3 - Added a Limit and Precaution for operation during a LOCA in order to complete with question 9A-49 of FSAR.
2.
Procedure B.3, " Normal Operation" Revision 1 - Updated procedure to incorporate latest Limits and Precau-tions and latest feed and bleed operating methods.
1.2.1.4 Surveillance Tests 1.
Auxiliary Feedwater bypass valve SFV-20578 failed to open on February 22, 1975 during performance of surveillance procedure SP 210.01C.
An Abnormal' Occurrence Report A0-75-6 was filed with NRC.
1.2.1.5 Periodic Containment Leak Rate Test There were no periodic containment leak rate tests performed in February, 1975 1.2.1.6 Changes, Tests and Experiments Authorized by NRC No changes, tests or experiments requiring authorization from Nuclear Regulatory Commission pursuant to 10 CFR 50.59a, were performed during February, 1975 e
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-l 1.2.2 POWER GENERATION Nuclear February 1975 Cumulative 1.
Number of hours reactor was critical 638 2,660 2.
Number of times reactor was made critical 3
16 3
Gross thermal power generated (MWH) 1,271,414 3,532,244 4.
Effective full power day.s 19.11 53 22 Electrical 1.
Gross electrical power generated (MWH) 432,792 1,141,991 2.
Net electrical power generated (MWH) 401,743 1,003,584 3
Numbar of hours generator was on line 598 2,338 e
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1.2.3 UNIT 1 SUUTDOWNS - February, 1975 t
i Status Corrective Action Duration Date Type
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Reason 2-2 Forced Power Reduced To tighten flanges dn Main Flanges tightened 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Reduc-from 92.6% FP Steam crossover piping.
tion to 20% FP r~}
2-5 Hanual Hot shutdown Repair Pressurizer Spray ' valve Valve repaired.
43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> I
which had failed open.
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Hot Shutdown Turbine-Generator tripped when Af ter thorough inspec-16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Gen /Rx an overspeed trip relay was tion, no failed Trip energized by either the SCOT components in SCOT or or EHC systems.
Subsequently, EHC systems could be i
Reactor tripped on high pressure found. New anti-during startup after the trip.
rotation pin was in-g No. I governor valve anti-stalled and valve rotation pin was noted to be was tested success-sheared.
fully.
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2-18 1x Trip Hot Shutdown Reactor tripped on low pressure ICS System taodified 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> during performance of load to eliminate problem.
Rejection Test from 75% FP.
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The low pressure trip was due to rapid cooldmen of the RCS when OTSG feedwater flow and subsequent heat sink became excessive during ICS runback.
2-23 Hanual Reactor ran Generator tripped as part of 39 minutes Gen.
back from 40%
Auxiliary Bus Transfer test.
1.2.3 UNIT l SHUTOOWNS - February, 1975 (Cont.)
Status Corrective Action Duration Date Type huta!
Reason
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2-20 Manual Reactor ran OCB's opened per Switchyard Hone required.
9 minutes Gen.
back from 25% blackout test.
Trip FP to 20% FP 2-23 Rx Trip Hot Shutdown Reactor tripped on Power Im-Pseudo design tran-.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> balance during performance of sient test rerun using design transient test.
a new control rod management schedule.
2-27 Forced Reduced Power Low Pressure Condenser Tube Condenser tube leaks 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Reduc-from 92% FP leak.
- repaired, tion to 55% FP a
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C-s February, 1975 1.2.4 MAINTENANCE OF SAFETY RELATED EQUlPMENT Description 1.2.4.1 Reactor Protection Pressure Trs,smitter Out of Calibration During steady state operation at 92% Full Power, the turbine generator tripped when an overspeed trip relay falsely became de-energized.
Reactor Coolant System pressure increased to 2355 psig tripping the reactor on reactor pro-tection high pressure Channels A, 8, and C.
Reactor protection high pressure Channel D did not trip during the transient. The D channel pressure trans-mitter was checked with the original calibration data and it was found that the zero reference had drifted. This drift increassd the pressure / response curve throughout the entira band resulting in an actual trip set point above the specified limit. 'The unit was removed from service and a unit of the same model was installed in its place. The new transmitter was calibrated and tested satisfactorily on the same day as the trip. The faulty transmitter was returned to the factory to determine the exact cause of failure. The failure of the reactor protection Channel D to trip did not cause any additional transient since Channels A, B, and C tripped the reactor.
Calibration verifi-cation of the remaining RPS pressure transmitters will be accomplished within 60 days.
- 1. 2. 4.'2 Auxiliary Feedwater Bypass Valve Failure Auxiliary feedwater bypass valve SFV-20578 failed to open during performance of surveillance procedure SP 210.01C. A local check of the valve and operator did not reveal any abnormalities. The valve was manually opened and closed from the local station.
During subsequent testing using the automatic operator, the valve was successfully opened and closed. The valve was automatically operated four times and the fourth operation was used to formally document com-pletion of the survelliance procedure. The valve will be retested monthly until the next required quarterly test for this valve.
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the valve operates properly, the surveillance testing will revert back to the
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normal schedule. The reactor was operating at 92% Full Power and no transient was experienced, due to this occurrence.
1.2.4.3 NI-RPS Count Rate Amplifier Out of Specs.
During performance of Source Range Channel test, it was found th'at a NI-RPS -
Source range count rate amplifier did not meet the required specifications.
The count rate ampilfier was replaced by a spare and returned to the manufact-urer for determination of cause and repair. The spare count rate amplifier was tested successfully prior to placing in service.
No plant transient occurred as a result of this problem.
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1.2 5 CHANGE lh PLANT OPERATING ORGANIZATION - FEBRUARY. 1975 There were no changes in the~ operating staff for those positions which are designated as key supervisory personnel on Fig. 6.2-1 of the Technical Spec-ifications.
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