ML19345G114
| ML19345G114 | |
| Person / Time | |
|---|---|
| Site: | Humboldt Bay |
| Issue date: | 11/26/1974 |
| From: | Searls F PACIFIC GAS & ELECTRIC CO. |
| To: | Goller K US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| 12121, NUDOCS 8103020477 | |
| Download: ML19345G114 (3) | |
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PACIFIC GAS AND E LE C T RIC C O M PANY FOWL 5 l
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,d Mr. Karl R. Goller Assistant Director for Operating ReactorsY%'"'" >
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Directorate of Licensing N/%Q' N
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Re:
Docket No. 50-133
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Dear Mr. Goller:
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i Pursuant to Section I.v,H.2.b. (2) of Appendix A heJt Humboldt Bay Unit No. 3 Facility Operating License we are her y
notifying you of two malfunctions.
The first malfunction involved the failure of the stem on one of the two main steam isolation valves; the second involved setpoint drift on one of the reactor trip accumu-lator control rod withdrawal low pressure interlock pressure switches.
Both of these malfunctions were discovered after the Unit was shutdown for its scheduled refueling and maintenance outage on October 30, 1974.
As a result of these events there was no radioactivity released to the environment, no unplanned personnel exposure, no injuries, no offsite consequences, nor loss or damage to the facility.
Main Steam Isolation Valve MO-6109 During a routine inspection on October 31, 1974 it was noted that the valve atem on the main steam isolation valve MO-6109 was broken.
The section of the stem attached to the motor operator was separated from the section in the valve by approxifnately 1 inch.
The break had occurred at a location not normally visible unless the valve packing is removed.
It was also noted that the main steam valve posi-tion switch actuating skiis had been bent when the stem over-traveled after breaking and would not have initiated a reactor trip until the valve was greater than 43% closed.
The failure apparently occurred on 1 "....?..i.
4 ppad7f P00R ORIGINAL.
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Mr. Karl R. Goller 2
November 26, 1974 October 28, 1974 when the valve was last operated as part of the monthly exercising of the automatic drywell isolation valves.
Exercising con-sists of a 25% closure of the valve followed by return to the backseat position.
A series of tests was performed which indicated that this malfunction did not cause or threaten to cause a degradation of the reactor containment or safety systems.
Two tests were conducted on the redundant main steam isolation valve MO-6101.
A reactor hydro-static test was conducted which verified that the main steam isolation containment boundary could be satisfactorily formed by only MO-6101.
The applicable routine operational tests required by Table VI-I of the Technical Specifications were performed which verified that MO-6101 would close automatically when required.
A third test was conducted which indicated that the bending of the reactor trip position switches also had no adverse effect on reactor safety.
These switches are de-signed to trip the reactor if the main steam isolation valve inadvert-ently closes.
The test indicated that it was not possible to close the broken valve using the motor operator nor was it possible for the valve to fall shut because it is mounted with the stem horizontal.
The valve might have closed slowly due to vibration but this would not be a situation requiring the rapid response of the reactor safety system.
Based on the above tests it is concluded that the malfunction of the main steam isclation valve MO-6109 did not degrade or threaten to de-grade the reactor containment and safety systems.
Inspection of the stem revealed that a crack about 20% of the stem's crosssectional area had occurred some time prior to the failure.
The stem is being tested further in an attempt to identify the cause of this cracking.
If the malfunction had not been discovered during a routine inspection it vould have been discovered during the routine monthly exercising of the valve.
There have been no similar failures of this type of valve.
The valve is a Chapman model 900# WEDS.
A new stem is being manufactured for MO-6109.
Strain gauges will be mounted on the new valve stem following assedbly of the valve and tests will be run to determine the forces the stem is subjected to during seating and backseating of the valve.
This test program will be satisfactorily completed prior to Unit startup.
Scram Accumulator Low Pressure Switch During the regularly scheduled testing of the control rod t
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Mr. Karl R. Goller 3
November 26, 1974 withdrawal permissive system it was discovered that the setpoint of one of the sixteen reactor trip accumulator low pressure switches had drifted down to 1298 psig, which is below the Technical Specifi-cation design valv.e of greater than 1300 psig.
The switch was in-spected and found to be in normal operating condition, reset to greater than 1300 psig and retested satisfactorily.
A review of previous test data was made to determine the expected drift range for this type of switch.
It was determined that the drift downward of this particular switch was within the expected range.
The control rod withdrawal permissive system operational test procedure is being revised to raise the setpoint of the trip accumulator pressure switches so that the setpoints will not fall below 1300 psig when normal setpoint drift is experienced.
The switch is a Teletron model 312-955 pressure switch.
Very truly yours,
/T CC:
Mr. R. H. Engelken, Director Directorate of Regulatory Operations Region V
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