ML20086L183
| ML20086L183 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/31/1973 |
| From: | Cooney M PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8402070450 | |
| Download: ML20086L183 (5) | |
Text
o o
\\
PHILADELPHIA ELECTRIC C6MPANY 2301 MARKET STREET i
a, PHILADELPHI A. PA.19101 (21518414ooo October 31, 1973 (V e
,OW fI Mr. A. Giambusso
[
\\
[/f NI(.b [!(pe h/g Deputy Dircctor of Reactor Projects United States Atomic Energy Commission D
- g. D$
'1 Directorate of Licensing g!.?> r7.g Washington, D. C.
20545
Dear Mr. Giambusso:
j n-
Reference:
Peach Bottom Atomic Power Station - Unit #2 6
7' Facility Operating License DPR-44 Docket No.
50-277
Subject:
Abnormal Occurrence During the week of October 22, 1973, the performance of routine surveillance tests disclosed three instrument deficiencies.
Also, a primary system relief valve bellows leak indication was noted. A radiation detector monitoring the refuel'ing floor ventila-tion exhaust duct showed a low reading during a calibration procedure; a Icvel switch which rnonitors the scram discharge volume level mal-functioned; and the flow biased input to an average power range monitor failed at an intermediate output level. These deficiencies were reported to Mr. R. A. Feil or Mr. Eldon Brunner, A.E.C. Region 1, Regulatory Operations Office.
In accordance with Section 6.7.2.A of the Technical Specifi-cation Appendix A of DPR-44 for Unit #2 Peach Bottom Atomic Power S ta ti on, these deficiencies are being reporte'd to the Directorate of Licensing as Abnormal Occurrences.
Refueling Floor Exhaust Radiation Detector A calibration procedure performed on the radiation detectors which moni tor the refueling floor and reactor building exhaust ducts disclosed that the calibration of radiation detector 4320 had shif ted such that the output of the detector was significantly less than the actual radiation. Based on the calibration data obtained, the detector l
would have had to be exposed to 24 mr/hr in order to trip. The Technical Specifications Table 3.2.0 requires the channel produce a trip output when radiation level exceeds 16 mr/hr.
hD 0
DR e
e tj COPY SENT ROI3d a
I
/~
(N
~
/
Mr. A. Gicmbu October 31, 1973 Page 2 1
Investigation and Corrective Action a-In order to identify the cause of the calibration shift, radiation monitor 4320 was bench calibrated.
This bench calibration indicated that the monitor and the detector were properly adjusted.
The detector was reinstalled and the field calibration ' repeated.
During the retest the radiation monitor tracked the reading on a calibrated GM instrument when exposed to an external radiation source. No defect in the instrument channel was noted that would cause the previous low output readings.
It is, therefore, concluded that the initial calibration data was in error and that no actual defect existed within the instrument channel.
Safety Implications The radiation detectors in the refueling floor ventilation exhaust duct initiate shutdown of the refueling floor ventilation system and initiate Standby Gas Treatment System operation on a correct one of two twice logic.
Since the investigation of the initial low output readings disclosed no channel deficiencies and a subsequent calibration showed the instrument channel to be functioning properly, it is assumed that the channel was in proper calibration at all times.
Since no instrument deficiency existed, there are no safety implica-I tions involved. This item was reported to the A.E.C. based on the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> reporting requirement before detailed investigation could be completed.
The results of the detailed investigation indicated that the item probably should not have been reported.
Scram Discharge Volume Level Switch Following the reactor scram on 10/22/73, a functional test on the scram discharge volume level switches was performed. This test disclosed that level switch 231D was defective.
Investigation An inspection of the level switch mechanism indicated that tbc switch pivots were adjusted too tightly and showed some signs of corrosion. The other three campanion switches were also inspected.
Level switch 231B also showed some signs of slight corrosion. The other two switches appeared to be in excellent condition.
Corrective Action The level swi tches 231B and D were cleaned, lubricated, adjusted and recalibrated. No signs of binding or improper operation were noted following this corrective action. Addi ti ona l l y, the 'D' icvel switch will be functionally tested on a' weekly basis until reliable performance is verified.
.]
o i
.o Mr. A. Giambusso October 31, 1973 Page 3 i
l l
Safety Implications t
The scram discharge volume high level switches are used in a one of two twice logic to scram the reactor when the water level in the scram discharge volume reaches 50 gallons. These switches ensure a re' actor shutdown before the available volume of the Scram Discharge Volume is reduced to the point where the reactor could not be scrammed.
With the reactor at power. this scram circuit would be activated if several control rods had a very high leakage fran the reactor through their scram exhaust valve to the header.
Presently, the leakage-from the reactor via this path is essentially zero.
Since 3 of the 4 level switches which monitor this function were operable, a sudden increase in leakage would have successfully produced the required reactor scram.
The failure of an individual switch in this logic circuit does not have significant safety implications.
A previous malf-: action of level switch 2-3-231D was reporte'd
~
to the AEC on 9/8/73. The cause of failure on the previous occasion was traced to a slight binding in the float mechani sm.
On this occasion the float mechani sm was free. The two failures appear to have been caused by independent problems.
During the performance of the surveillance test on the APRM channels prior to placing the reactor in the RUN mode, the flow biased circuit of channel F was found to be defective.
The flow biased output had failed with an output representative of approximately 30% recircula-tion flow. Reactor flow at this point and time was 30%. The reactor was operating with a power level of approximately 1%.
Investigation To determihe the cause of failure the flow biased network was bench tested to locate several bad Zener diodes in the flow control trip reference card.
Corrective Action A new card was substituted in channel F, calibrated, and the surveillance test repeated.
Testing performed on the same day on the other 5 APRM channels indicated that all other channels were fully opera-tional.
The failure of these components is believed to be an isolated even t.
-9W g
O e= > *
~
Mr. A. Git so Octcbar 31, 973 e
Safety Implications At the time that the surveillance test was performed
- the reactor mode swi tch was in the STARTUP post tion. This bypasses the flow biased APRM scram function. The surveillance test simulates the mode switch in RUN in order to check out the flow biased scram circuits prior to establishing a power level where these functions become activated.
Since the flow biased circui t failed with approxi-mately a 30% output, the trip point from channel F would have been conservative with recirculation flows greater than 30%. Since the recirculation flow at the time of failure was 30%, this failure is believed to be in the safe direction. The surveillance test per-formed on 10/23/73 also indicated that the non-flow biased trips set at less than 12% for Rod Block and less than 15% for Scram, were operable wi th the proper setpoints.
Since 5 of the 6 flow biased scram circuits were operable, a scram would have occurred at the proper setpoint under any recirculation flow condition with the re-actor mode switch in "RUN".
No safety implications are attached to the above failure.
l Peactor Relief Valve Bellows Failure Indic'ations i
On10/23/73, the circuit which monitors the integrity of the bellows on relief valve 2-71K indicated a. bellows leak. The indication was initially suspected as being an instrumentation problem but was recogni:cd by the PORC as a possit>1e relief valve bellows..
fai lu re.
In accordance with paragraph 3.6.D.2.(a) of the Technical Specifications the PORC determined that reactor operation could con-tinue for 30 days Ghile investigation continues.
Investigation The' electrical connections at the relief valve pressure switch were inspected on 10/26/73 and found to be in good condition.
1.ater on the same day the reactor pressure was reduced to less than 150 psig.
When the pressure dropped below the setpoint of the pressure switch, the bellows failure indicating light extinguished, indicating that the bellows or bellows associated "0" rings have developed a leak.
Corrective Action The leaky bellows or "0" rings on this relief valve will be replaced within the 30 day period either by replacing the defective component, or replacing the enti re relief valve.
If this cannot be accomplished, the reactor will be p; aced in a cold shutdown condition prior to November 23, 1973.
m-
['
s.
~
s l
Mr. A. Giambusso
~
October 31, 1973 Page 5 4
Safety Implications The Unit 2 reactor is presently in the startup testing phase until vessel testing and testing at 150 psig has been completed. The power level of the reactor is not expected to exceed 50% prior to 11/23/73. With the reduced reactor power, the number of relief valves requi red is signi ficantly reduced.
Thi s fai lure, therefore, has no safety implication.
Very truly yourh,
/L-(w6
. J. Cooney Asst. Gent. Supt.'
e l
Generation Division I
cc.
Mr. J. P. O'Reilly l
Director, Region 1 i
United States Atomic Energy Commission 631 Park Avenue King of Prussia, PA 19406
'