ML17331A664
| ML17331A664 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/10/1981 |
| From: | Shaller D INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA) |
| Shared Package | |
| ML17331A665 | List: |
| References | |
| NUDOCS 8103100342 | |
| Download: ML17331A664 (19) | |
Text
~+I REGULATO NFORMATION DISTRIBUTION 'EM (RIDS)
ACCESSION NBR:8103100302 DOC ~ DATE! 81/02/10 NOTARIZEDI=NO, DOCKET FACIL:50 315 Donald C.
Cook Nuclear Power Planti Unit ii Indiana 8
05000315 AUTH',NAME AUTHOR AFF ILIATION SHALLERED ~ VS Indiana 8 Michigan Electric Co ~
RECIPSNAME RECIPIENT AFFILIATION Office of Management and Program Analysis
SUBJECT:
Forwards monthly operating rept for Jan
- 1981, DISTRIBUTION CODE:
AOOSS COPIES RECEIVED:LTR ENCL SI2E" T:LITLE: Annualr Semi Annual 0 Monthly OPerating RePorts-(OL Stage)
"-NOTESr ISE:3 copies all material.
05000315 EXTERNAL: ACRS LPDR 11 03 RECIPIENT ID CODE/NAME ACTION:
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~yucca INDIANA& MICHIGANFlECTRIC COMPANY DONALDC. COOK NUCLEAR PLANT P.O. Box 468, Bridgman, Michigan 49106 (616) 466.6901 February 10, 1981 Director, Office of Management Information and Program Control U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Pursuant to the requirements of Donald C.
Cook Nuclear Plant Unit 1
Technical Specification 6.9,1.6, the attached Monthly Operating Report for the month of January, 1981 is submitted.
Sincerely, D. V.
Shaller'lant Manager
DVS:ab Attachments IP QJs CXC CC:
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F. Kroeger H. L. Sobel J.
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OAVERAGE OAILY UNIT PGNER LEVELS OGCKET NO.
UNIT 1
DATE 2-5-81 COMPLETED BYA. L. Tetzlaff TELEPHONE (616) 465-5901 MONTH Januar 1981 DAY AVERAGE DAILY POWER LEVEL (MWE-Net)
DAY AVERAGE DAILY POWER LEVEL (MWe-Net) 10 12 13 15 16 325 1052 1
5 1052 17 18 20 21 22 23 24 25 26 27 29 30 1037 1052 1
5 INSTRUCTIONS On this format list the average daily unit power level in NWe-Net for each day in the reporting month.
Compute to the nearest whole megawatt.
UNITSIIUTI)01VNS ANI)POlYER REI)UCTIONS REPORT h(ONTII january, 1981 I)OCKET NO.
COh(PLETEO I)y B.A. Svensson TELEPIIONE 616 465-5901
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CJ Ca n sc A. Co fIcc Ilvc A<<lion tll Prcvcnl Rccnfrcnce I 69 Cont'd 170 810107 F
18.7 N.A.
HA INSTRU u age con snue rom previous mon The unit was returned to service on 810105.
Total length of outage 294.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
Reactor power increased to 100/. 810107.
Unit trip due to safety injection actuation/reactor trip.
Safety in-jection was due to high differential pressure between steam leads caused
'y low pressure in No..3 steam gen-erator.
The initiating cause was the No.
2 main turbine stop valve drifting off its wide open position due to a faulty servo valve.
This caused the other three main stop valves to go closed resulting in a sudden increase in steam flow from No.
3 steam gen-erator thru No.
2 turbine stop valve with the resulting drop in No.
3 steam generator pressure.
Unit was I
I': Fo(<<cd S: Scllcdnicd
(')/77)
Reason:
A-I:linipnlcnt Failnrc (Explain)
I).h(aintcnancc or Test C-Refnclinl; I) Rcgolalt oly Restriction I.'-Operator 'I'rainin(,8c Liccnsc I'.xalninallon F-Administrative G Oi)era tlonal L'Iror (I'.xplain)
I I Other (Explabl) 3 hlelhod:
I -hlanllal 2 Manual Scranl.
3.Aulonla tie Scraln.
0 Otllcf(Explain) 1005 power4the same day.
Exhibit G - Instroctions for Preparation ol l)ata LollySllccls for Li<<cnscc Evdnt Report (I.I!R)File (NURI.'G.
OI6 I )
5
'I'xllibilI - Salnc Source
NITSHUTDOWNS AND POWER REDU S
INSTRUCTIONS This report should describe all plant shutdowns during the report period.
In addition. it should be the source of explan-ation of significant dips in average power levels.
Each signi-ficant reduction in power level (greater than 20% reduction in average daily power level for the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) should be noted, even though the unit may not have been shut down completelyl.
For such reductions in power level, the duration should be listed as zero, the method of reduction should be listed as 4 (Other), and the Cause and Corrective Action to Prevent Recurrencc column should explain.
The Cause and Corrective Action to Prevent Recurrence column should be used to provide any needed explanation to fully describe the circumstances of the outage or power reduction.
NUMBER.
This column should indicate the sequential num-ber assigned to each shutdown or significant reduction in power for that calendar year.
When a shutdown or significant power reduction begins in one report period and ends in another.
an entry should bc made for both report periods to be sure all shutdowns or significant power reductions are reported.
Until a unit has achieved its first power generation, no num-ber shouid be assigned to each entry.
DATE.
This column should indicate the date of the start of each shutdown or signiricant power reduction.
Report as year. month. and day.
Auaust i4. 1977 would be reported as 770814.
When a shutdown or significant power reduction begins in one report period and ends in another, an entry should be made for both report periods to be sure ail shutdowns or,significant power reducuons are reported.
TYPE., Use "F" or "S" to indicate either "Forced" or "Sche-duled," respectively, for each shutdown or significant power reduction.
Forced shutdowns include those required to be initiated-by no later than the weekend following discovery of an off-normal condition.
It is recognized that some judg-ment is required in categorizing shutdowns in this way.
In general.
a forced shutdown is one that would not have been completed in the absence of the condition for which corrective action was taken.
DURATION.
Seifwxplanatory.
When a shutdown extends beyond the end of a rcport period, count only the time to the end of the report period'and pick up the ensuing down time in the following report periods.
Report duratibn of outages rounded tu the nearest tenth ofan hour to facilitate summation.
The sum of the total outage hours pius the hours the genera-tor was on line should equal the gross hours in the reporting period.
REASON.
Categorize by letter designatiun in accordance with the table appearing un the report form. -Ifcategory H must be used. supply brief comments.
METHOD OF SHUTTING DOWN THE REACTOR OR REDUCING POWER.
Categorize by number designation Nuic that this differs i'rom the Edison Electric Institute (EEI) definitions of -Forced Partial Outage-and
-Sche.
duied Partial Outage.-
Fur these turin~. FFI uses a change ui'0 MW as the break point.
Fur larger puv er reactors. SO MW is Iuu snlail a change iu warrant cxplanatiuu.
in accordance with the table appearing on the report form.
Ifcategory 4 inust be used, supply brief comments.
LICENSEE EVENT REPORT
='.
Refcrenc.
thc applicable reportable occurrence pertaining to the outage or power reduction.
Enter the first four parts (event year. sequential report number, occurrence code and reoort type) of the five part designation as described in Item 17 of Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG416l).
This information may not be immediately evident for all such shutdowns, of course, since further investigation may be required to ascertain whether ur not a reportable occurrence was involved.) If the outage or power reduction will not result in a reportable occurrence.
the positive indication of this lack of correhtion should be noted as not applicable (N/A).
SYSTFM CODE.
Thc system in which the outage or power reduction originated should be noted by the two digit code o t Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG4161).
Systems that do not fit any existing code should be designa-ted XX. The code ZZ should bc used for those events where a system is not applicable.
COMPONE'iiT CODE.
Select the most appropriate component from Exhibit I
~ Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG4161).
using the followingcritieria:
A. Ifa component failed, use the component direcriy involved.
B. If not a component failure, use the related component:
eg..
wrong valve operated through error: list valve as component.
C.
If a chain of failures occurs, the first component to mai.
function should be listed. The sequence of events. includ-ing the other components which fail, should be described under the Cause and Correcuve Action to Prevent Recur-rence column.
Components that do not fit any existing code should be de-signated XXXXXX. The, code ZZZZZZ should be used for events where a
component designation is not applicablc.
CAUSE L, CORRECTIVE ACTION TO PREVENT RECUR-RENCE.
Use the column in a narrative fashion to ampiify or explain the circumstanc s of the shutdown or power reduction.
The column should include the specific cause for each shut ~
down or significant power reduction and the immediate and contemplated long term corrective action taken. ifappropri-ate.
Ttus column should also be used for a description of the major safety.rehted uorrective maintenance performed during the outage or power reduction including an identification of the critical path activity and a report of any single reiease of radioactivity vr single radiation exposure spcciri<<ally associ
~
ated with the outage which accounts for more than 10 percent ui the allowable annual values.
For long textual reports continue narrative on separate naper and rcicrence the shutdown or power reduciiun i'vr this ilaffative.
Doc et No.:
Unit Name:
Completed By:
Telephone:
Date:
Page:
50-315 D.
C.
Cook Unit 81 D.
R. Campbell (616) 465-5901 February 10, 1981 1of2 MONTHLY OPERATING ACTIVITIES - JANUARY, 1981 The Unit entered the reporting period in Mode 5 in a.planned outage for general maintance and inspection of the ice condensor doors.
Heatup of the Unit began January;4, 1981..
The Unit entered Mode 4 at 1204 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.58122e-4 months <br /> and Mode 3 at 1808 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.87944e-4 months <br />.
The i;eactor was made critical at 1622 hours0.0188 days <br />0.451 hours <br />0.00268 weeks <br />6.17171e-4 months <br />, January 5,
1981 and entered Mode 1 at 1733 hours0.0201 days <br />0.481 hours <br />0.00287 weeks <br />6.594065e-4 months <br /> the same day.
The reactor reached 100%
power at 0858 hours0.00993 days <br />0.238 hours <br />0.00142 weeks <br />3.26469e-4 months <br />, January 7, 1981.
On January 7, 1981 the Unit experienced a reactor trip with full safety injection.
All ECCS systems operated as designed.
See summary for details.
The reactor was made critical at 2245 hours0.026 days <br />0.624 hours <br />0.00371 weeks <br />8.542225e-4 months <br />, January 7, 1981.
The Unit reached 100/ power at 2205 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.390025e-4 months <br />, January 8, 1981.
The Unit operated the remainder of this reporting period at 90 to 100/ reactor power.
\\
Total electrical generation for the month was 631,790 mwh.-
1/7/81 At 0922 hours0.0107 days <br />0.256 hours <br />0.00152 weeks <br />3.50821e-4 months <br />, the Unit experienced a reactor trip and safety injection.
The safety injection was actuated due to high differential pressure between steam leads, caused by a low pressure in number 3
The number 2 main turbine stop valve drifted off the open position due to a faulty servo valve.
The other three stop valves are slaved off of the open limit switch, hence upon its drop out, these three valves went closed by means of their slow close-solenoids.
There is a cross tie between the four steam lines upstream of the stop valves, but the connection from each steam lead to the cross tie is limited in capacity by an orifice.
The only stop valve thus remaining open was the number 2 stop valve, resulting in a sudden increase in steam flow in this liri'e.
As 'the three stop valves were closing the safety injection occurred when the differential pressure between the steam. line, feeding number 2 stop valve (from number 13 steam generator) and the other steam.lines reached the 100 PSID set point, the reactor trip and safety injection was initiated.
The defective servo valve was replaced.
III
Docket No.:
Unit Name:
Completed By:
Telephone:
"'Date:
Page:
50-315 D.'-C:. Cook Unit 8l D.
R. Campbell (616) 465-5901 February 10, 1981 20f2 Summary (continued):
1/8/81 R-26 alarmed, high, the cause was determined to be a
leaking root val.ve CS374 located on the VCT sample line.
The valve has been repaired and tested for leaks.
1/9/81 The East motor, driven Auxiliary Feed Pump was inoperable to repair emergency suction supply valve from the Essential Service Water system.
1/12/81 The B.I.T. and associated B.A.S.T. were out of Tech.
Spec'.
when they were diluted to 19,300 PPM.
The shutdown of the Unit was started and stopped when the B. I.T. and B.A.S.T.
was brought back in specification.
1/9/81 The Turbine driven Auxiliary Feedwater Pump was out of service to repair drains.
1/20/81 CD Emergency Diesel Generator was out of service to replace a diaphram in the starting air pilot valve.
We had rod urgent failure alarms on three seperate occasions on 1/21/81 and 1/22/81 the cause was loose logic cards and a blown fuse.
DOCKET NO.
UNIT NAME DATE COMPLETED BY TELEPHONE PAGE 50 - 315 D.
C'.
Cook - Unit No.
1 2-10-81 B. A. Svensson 616 465-5901 1 of!
1 MAJOR SAFETY-RELATED MAINTENANCE JANUARY 1981 M-2 M-3 M-4 BD-.101-4, shell drain valve on blowdown line from No. 4 steam gen-erator had a body-to-bonnet leak and was found to be steam cut.
The valve was replaced.
4 Seal water orifice flange of No.
4 RC pump leaked.
Replaced gas-kets in both flanges.
Charging line high point vent valve, CS-345, had body-to-bonnet leak.
Replaced packing cartridge and bonnet-seat assembly.
Inspected pressurizer spray valves for packing and body-bonnet leaks.
Replaced bonnet gasket on NRV-164 and repacked NRV-163 and NRV-164.
Valves tested satisfactorily.
Flange on Loop 2 RTD header vent was leaking.
Replaced flex gas-ket.
M-7 Flange at flow orifice NFA-240 on loop 4 RTD, header return to pump suction leaked.
Replaced flex gasket.
The handholes of all four steam generators were removed and the tubes in the area of the tube lane blocking devices were inspected.
Some tube damage was detected in steam generators 1
and 4.
The affected tubes were plugged.
The containment isolation check valve on the CCll supply to the cooling coil for the main steam penetration, CCll-243-72, was leaking by.
The check valve was disassembled, cleaned and re-assembled.
M-10 Pressurizer power operated relief valve, NRV-153, was leaking by.
Replaced the plug, stem, gaskets and packing.
Also remachined the seats.
Valve retested satisfactorily.
Pressurizer power operated relief valves, NRV-151 and 152 were leaking by.
Gaskets and packing were replaced on NRV-151 and the plug and seats were remachined.
Replaced the plug, stem, packing and gaskets on NRV-152 and remachined the seat.
Both valves were retested satisfactorily.
e i $'l 1
DOCKETNO.
UNIT'AME DATE COMPLETED BY TELEPHONE PAGE 50 - 315
'D.
C.
Cook - Unit No.
1 2-10-81 B. A. Svensson 616 465-5901 2 of 4'AJOR SAFETY-RELATED MAINTENANCE JANUARY, 1981 Charging line motor operated isolation valve, gM0-200, would not open.
Removed the valve operator and freed the valve, Reinstalled the operator, replaced the motor and repaired the operator.
Reset the operator limits and tested the valve.
The 2CD2 diesel jacket water pump discharge check valve, DG-153C, would not seat.
Replaced the spring and lapped the disc.
Had valve tested.
The component cooling water supply check valve to the RCP motors, CCW-122, appeared to be stuck open.
Inspected valve and lapped discs'eassembled valve and it functioned properly.
The West RHR pump was experiencing intermittent losses of control power.
Inspected and found a fuse clip on the main fuses loose.
Adjusted clip, cycled pump and no further problems were experienced.
Trip and throttle valve for the auxiliary feedpump turbine would not close.
Cleaned the contacts of the torque switch.
Valve tested satisfactorily.
The cooling water line to the oil cooler for the turbine driven auxiliary feedpump had a leak at a flanged joint.
The motor operated ESW isolation valve to the east motor driven auxiliary feedpump, WM0-754, leaked by.
Replaced the valve and the motor on the limitorque operator.
The south safety injection pump seal water line was leaking.
Re-moved pump from service.
Replaced a nut and ferrule on swagelok fitting and returned pump to service.
The volume control tank vent regulating valve, RRV-300, had a
Repacked the valve and had it tested.
The instrument line downstream of the root valve for gFI-200, charging header flow instrument, was leaking.
There was a crack in the weld where the 3/8" tubing entered a reducing bushing, Ground out old weld, cut off tubing and rewelded.
Had the re-quired NDE performed.
5
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DOCKET NO.
UNIT NAME DATE COMPLETED BY TELEPHONE PAGE 50 - 315 D.
C.
Cook - Unit No.
1 2-10-8 B. A. Svensson 616 465-5901 3 of 4 MAJOR SAFETY-RELATED MAINTENANCE JANUARY, 1981 The north waste gas compressor inlet valve, CCM-151N, had a bent stem.
Replaced the stem and valve bonnet.
Flange downstream of the reciprocating charging pump was leaking.
IRV-310, the east residual heat exchanger discharge valve, failed to the open position.
An air line fitting was found broken.
A new fitting was installed and proper operation of the valve was verified.
The air hoses on NRV-151, NRV-152 and NRV-153, pressurizer power operated relief valves, were found cracked and the valve closure time was approaching the ISI time limit of 5 seconds.
The hoses were replaced with larger hoses and the closure time of the valve wa,s reduced to less than 2.5 seconds on all valves.
The one inch safety valve, SV-141, downstream of FRV-275, east motor driven auxiliary feedpump emerqency leak-off valve, was found leaking when FRV-257 was opened.
The safety valve had been damaged due to the setpoint being too low.
The valve was repaired and the setpoint raised to 1500 psig.
HARV-040, No.
4 reactor coolant pump No.
1 seal leak-off valve, was closed during a surveillance test and could not be reopened.
The solenoid for gRV-040, XS0-28, located in the regeneration heat ex-changer room was removed and a spare installed.
Also, the solenoid for HARV-160 and HARV-161 was replaced at this time.
Problems were previously encountered on HARV-161 solenoid.
The solenoid model was HT 831654, 250 VDC.
Turbine stop valve No.
2 developed a drift problem tending to close.
The valve's servo amplifier was replaced and stop valves No.
1 and No. 2 were checked for proper operation.
I On January 7, 1981, when Unit 1'ripped from 100/ "power, the events sequence monitor indicated No.
1 stop valve closure was the first event in the trip.
Upon opening the stop valves on a depressurized steam header, all valves appeared to function properly.
Each valve was tested in sequence with no abnormalities in operation observed.
No.
1 stop valve showed no binding in opening or closing, nor was any drift observed from the open position.
/ Continued...........
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DOCKET NO.
UNIT NAME DATE COMPLETED BY TELEPHONE PAGE 50 - 315 D.
C.
Cook - Unit No.
1 2-10-81 B. A. Svensson 616 465-5901 4of4 MAJOR SAFETY-RELATED MAINTENANCE JANUARY, 1981 C&I-5 Cont d.
A G.
E.
Company representative was called on-.site for further checkout of the stop valves.
All limit switches were tested (with the header depressurized) and no problems were found.
After Unit 1 was critical and the steam lines were pressurized, the valves were again tested.
After 3 cycles, No.
2 stop valve drifted slightly closed, not enough to clear the open limit
, switch, but the drift was visible.
C&I-6 C&I-7 C&I-8 C&I-9 On each successive
- cycle, No.
2 valve drifted further closed.
At the suggestion of the G.
E. representative, the servo amplifier on No.
2 valve was replaced and proper operation of the valve was observed for 5 cycles, after which it was returned to service.
An urgent failure alarm on the rod control system Group 2 of con-trol bank D was received.
The problem was traced to a blown fuse in Phase C of the moveable gripper input power supply.
The fuse was replaced with a spare.
Proper system operation was verified and the rod groups were realigned.
Radiation monitoring system channels R25 and R26'ndicated a maxi-mum flow of approximately 9 CFM.
The flow meter was replaced with a spare.
The spare flow meter indicated 15 CFM.
The monitor was
,adjusted for correct flow following the functional check of the high and'ow'flowtrips.'I Investigate problem with monitor lights on ECCS panel being very dim.
Of the two power supplies auctioneered supplying 24 vdc to the light circuits, one power supply was found tripped and could not be reset.
The second power supply output voltage was low and could not be increased through adjustment.
Silicon controlled
,rectifiers (SCR's) were replaced in the failed power supply which restored the power supply to an operable status.
The second power supply components required replacement.
The second power supply was adjusted to provide the correct output voltage.
QLS-951, primary water storage tank level transmitter, failed to the low end of scale.
The temporary heat tape on the sensing line caused the water in the sensing line to boil, resulting in the channel failure.
The heat tape was deenergized.
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