ML20030A490
| ML20030A490 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/22/1969 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090747 | |
| Download: ML20030A490 (12) | |
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DEC 221969 - 1 g
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-2 CONSUMEBS POWER COMPANY
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Docket No 50-155 0)
I Report of Operation of Big Rock Point Nuclear Plantg,vp+-
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License No DPR-6 Ret!M rt'r M,g,47 May 1, 1969 Through October 31, 1969
'I.
SUMMARY
OF OPERATIONS A.
Power Operation The plant was shut down at the beginning of the report period with the sixth refueling outage in progress. Members of the Ccmpany-vide union (Utility Workers Union of America) were still on strike.
(See Tenth Semiannual Report.) The refueling operations (including fuel loading) were completed by supervisory personnel. Other neces-sary maintenance work was performed by supervisory personnel, engineers and technicians. No significant problems were encountered.
The plant resumed operation on May 9, 1969 generating 53 MWe(g).
Off-gas activity was 1,600 pCi/sec.
Supervisory personnel, engineers and technicians continued to operate the plant until June 30, 1969 when the strike was settled and union personnel returned to work.
Due to the premature failure of several "E" fuel bundles (see Tenth Semiannual Report), the plant was operated at 53 MWe(g) during this report period to conserve reactivity and to reduce heat flux on the fuel cladding in the core, thereby possibly reducing fuel cladding deterioration until an analysis can be made to determine the mechanism of fuel failure. Water chemistry tests are now being performed in con-I
, junction with General Electric Company (GE) to study crud effect prob-lems in respect to the recent fuel failures (see Test Section V-B).
There were two short outages in the month of June 1969 On June 7, the unit was forced out of service to repack the outside gland on the butterfly valve in the No 1 reactor recirculation loop.
(See Mainte-i nance Section IV-C.) The un.~.t was removed from service on a scheduled out-age June 21. 1969 to repair four steam leaks in the turbine rire tunnel area.
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2 Off-gas activity increased to approximately 2,000 pCi/sec in early July 1969 and continued on an upward trend to 2,500 pci/see near the end of the month. On July 7, the unit was removed from service on a scheduled outage to repair leaks in the turbine stage drains and in the B-3 control rod drive cooling water connections.
On August 11, 1969, the unit van removed from service on a scheduled outage to repair various stear
-uks, and to inspect for known leakage in the turbine main condenser and core spray heat exchanger.
(See Maintenance Section IV-E and G.) The unit was returned to service on August 13, and operated continuously for 62 days.
On September 27, the inner seal on the No 2 recirculating pump partially failed. However, plant operation was unaffected as the outer seal continued to perform satisfactorily. Seal temperatures were within operating limits. There was a temporary increase in the off-gas activity on September 29, 1969 from approximately 2,800 to 3,400 pCi/sec. However, the activity gradually returned to approximately 2,800 pCi/sec.
The plant was removed from service on a scheduled outage October 16, 1969 to perform semiannual control rod drive testing and to repair several steam leaks. A subsequent inspection revealed a leak on No 1 recirculation pump suction valve which necessitated replacement of the Flexitallic bonnet gasket. The main steam bypass valve was repacked and a new packing gland installed during this outage. Start-up was de-layed by failure of the poison system Valve CV-4050 to open.
(See Main-tenance Section IV-K.)
The unit was returned to service on October 20, 1969 Shortly after start-up on October 20, the unit was forced out of service due to high conductivity of the primary coolant.
Subsequent investigation revealed that during reactor water blowdown to radvaste, some of the water had been drawn through the clean-up demineralizer with the nonregenerative heat exchanger valved for reduced cooling water flow.
The temperature limit of the resins was exceeded resulting in a chemical breakdown of the resins. The demineralizer resin bed was replaced and I
when the reactor water conductivity returned to within Technical Speci-fication limits, the plant was started up.
The unit was placed on the line setober 21.
Following this start-up, the off-gas activity peaked
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9 to approximately h,000 pCi/see and had leveled off to approximately 3,h00 upci/see by the end of the period.
B.
Refueling Outage Various items of interest which occurred during the sixth re-4 fueling vere. presented in our Tenth Semiannual Report.
The new core loading, after-completing the sixth refueling outage, consisted of the following:
- 1. ' Developmental Bundle 1
- 2.. Reload "B" Bundles 14 3.
Reload "C" Bundles 8
3 h.
Reload "E" Bundles 61 The Canadian Westinghouse fuel bundle arrived on site June 25, and will be inserted during the seventh refueling outage tentatively sched-I' uled for February 1970. This bundle was originally sch'eduled to be in-serted during the sixth refueling outage.
C.
Statistics 4
The reactor was brought critical nine times during the report period. The reactor was critical for 3,985 hours0.0114 days <br />0.274 hours <br />0.00163 weeks <br />3.747925e-4 months <br /> with electrical gen-eration of 208,089 MWh (gross) or 198,188 MWh (net). The thermal. output of the reactor' vas 644,748 MWh.
j II.
ROUTINE RELEASES. DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIAL h
I A.
A total of approximately 3.6 x 10 curies of activation and fission gases was released to the environs via the stack. This figure is based 3
upon 3,985 hours0.0114 days <br />0.274 hours <br />0.00163 weeks <br />3.747925e-4 months <br /> critical at an average release rate of 2.5 x 10 uCi/sec.
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B.
During this reporting period, the liquid radioactivity, released to Lake Michigan by way of the circulating water discharge canal, numbered 33 batches, with a total activity of 10.1 curies. All batches were re-leased on an identified basis by isotopic composition showing that approxi-mately h3 percent of the' activity was 2n-65. The bulk of the remaining-activity was-I-131, Cs-137, Cs-134 ar.d Co-58.
i C.
A total of 21 off-site shipments of radioactive material was made
'during this reporting period as follows:
Transfer Transfer j
Shipment No Date From To Radioactive Material
'l 5/ 1/69 DPR-6 GE-Val, 0017-60 3 Sets of "0" Rings From (Calif)
Rod Drives - 0.5 mci j-2
-5/ 2/69 DPR GE-Val, 0017-60 Beactor Water Filtrate and (Calif)
Crud - 116 mci i
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k Shipment Transfer Transfer No Date From To Radioactive Material 3
5/ 9/69 DPR-6 GE-Val, 0017-60 Irradiated Meta 111e Speci-DOT SP-5971 mens - 1690 Ci h
5/ 9/69 DPR-6 GE-Val, 0017-60 Profilmeter, Toole -
(Calif) 2 mci 5
5/25/69 DPR-6 GE-Val, 0017-60 Feed Water Crud and Fil-(Calif) trate 6 mci 6
6/12/69 DPR-6 GE-Val, 0017-60 GE Apparatus From Heater (Calif)
Drain Tests - 1 mci 7
6/23/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 0.2 mci 8
7/ 9/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 2 mci 9
T/22/69 DPR-6 NPI, 19-12667-01, Irradiated Cobalt - 450,000 DOT SP-535h Ci 10 7/30/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 1.5 mci 11 8/ 5/69 DPR-6 NRL 8-1393-2 Irradiated Reactor vessel 4
(A-66) BE Specimens - 13.9 Ci Permit 727 12 8/22/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 0.h mci 13 9/10/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fi1-(Calif) trate - 1.5 mci lh 9/11/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 0.1 mci 15 9/26/69 DPR-6 Isotopes Inc 200 ml Reactor Water; 29-55-6 200 ml Condensate - 0.1 mci 16 10/ 2/69 DPR-6 University of 1 Gallon of Reector Water -
Michigan,21-215 h 2 mci 17 10/ 9/69 DPR-6 Nuclear Engineering High-Level Radwaste 110 16-NSF-1, Amend #10 Ft3 - 23.6 C1 36 10/31/69 DPR-6 Nuclear Engineering 86-55 Gallon Barrels of 16-NSF-1, Amend #10 Waste - 257 mci 19 10/lk/69 DPR-6 Nuclear Engineering High-Level Radvaste 110 16-NSF-1, Amend #10 Ft3 - 37.5 Ci 20 10/13/69 DPR-6 GE-Val, 0017-60 Feed-Water Crud and Fil-(Calif) trate - 1 mci
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21 10/30/69 DPR-6 GE-Val, 0017-60 Reactor Water Crud - 250 (Calif) mci l
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- p III. RADI0 ACTIVITY LEVELS IN PRINCIPLE FLUID SYSTEM (FOR SIX MONTHS)
A.
Primary Coolant Minimum Average Maximum j
Reactor Water Filtrate "
-1
-1
~ uCi/cc.
1.h5 x 10 h.36 x 10 7.29 x 10
-Reactor Water Crud *
-1
-1 1.h5 x 10-1.31 x 10 5.8 x 10 pCi/cc/ Turbidity )
Iodine Activity (b
-2 4
4 pCi/cc 8'x 10 1.1 x 10 3 x 10 B.
Reactor Cooling Water System Minimum Averace Maximum Reactor Cooling Water "
-3 pCi/cc 4.36 x 10 l.3 x 10 2.92 x 10 ^
The principal radionuclides in the reactor cooling water system were K h2 and Cr-51. These radionuclides resulted from the activation of the potassium chromate inhibitor.
C.
Spent Fuel Pool - Radioactivity in the spent fuel pool is prin-
'cipally activated corrosion products from stored fuel and core components.
j.
Minimum Averace Maximum Fuel Storage Pool (a)
-3
-2
-2 pCi/cc 2.9 x 10 1.31 x 10 8.7 x 10 Fuel Pool Iodine
-6 h
pCi/cc 2 x 10 1 x 10 9.5 x 10-3 IV.
PRINCIPAL MAINTENANCE PERFORMED A.
The adjustable split vedge support installed under the outside equipment lock of the containment sphere (see tenth semiannual report) has been adjusted to allow a 1/8-inch clearance between bearing surfaces.
Bechtel Corporation and Chicago Bridge and Iron Company determined the
" A counter efficiency based on a gamma energy of 0.662 Mev and one gamma photon per disintegration. Decay scheme is assumed to convert count rate to microcuries. All count rates were taken at two hours after sampling.
(b) Based on efficiency of Iodine-131, two hours after sampling.
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- Based on APHA turbidity units and 500 ml of filtered sample.
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vedge was unnecessary to support the loaded spent fuel shipping cask moving through the equipment lock and recommended 1/8-inch clearance be maintained to accommodate the maximum vertical movement obtainable from design temperature extremes on the containment sphere.
B.
Control rod drives in Positions C-2 and C h were replaced with rebuilt spares. These replacements resulted from our continuing overall preventative maintenance program. Control rod Drives B-3 and D-6 flanges were found to be leaking.
"O" rings were replaced in each of these flanges (see Tenth Semiannual Report for details on-similar type leaks).
I C.
A steam leak developed from packing on the outside gland on the i
' butterfly valve in No 1 reactor recirculation loop causing a plant out-age. This valve was successfully repacked.
D.
Inspection revealed a leak on the No 1 reactor recirculation pump suction valve which resulted in the replacing of the Flexitallic gasket in the bonnet. During repairs, indication of a small leak was observed on-the No 2 pump suction valve and the bolts were torqued.
I Both valve packings were tightened and hydrostatically tested with no leakage observed.
E.
The turbine main condenser was checked for tube leaks with one j
leaking tube being found and plugged. No additional tube leaks were noted. The condenser tubes are made of Admirality metal. This was the first tube leak detected since initial plant operation.
F.
The No 2 plant exhaust fan was disassembled, both upper and lover bearing assemblies replaced and the fan returned to service. 7 was the second bearing replacement since initial plant operation.
G.
The tube side end cover was removed from the core spray heat ex.
changer for inspection. A small leak was found in one tube and plugged.
No other indication of leaking tubes was observed. The "U"-tubes in this heat exchanger are made of low carbon steel. This was the first tube leak detected since initial plant operation.
H.
Overheating of the diesel generator was experienced during a test run. The cooling water pump suction piping check valve was found to be plugged with a small fish. The heat exchanger and suction line check valve were cleaned immediately and strainer bars were temporarily installed at
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the open end of this line. The generator was restarted and performed satisfactorily. A well point was' installed at a later date on the I
cooling water suction line to prevent this from happening again, cI.
New eylinder head ' gaskets were installed in the No 1 control rod drive pump and all three cylinders were repacked. The pump was previously repacked in May 1968. This represents 16 months of operation and was the first time stainless' steel nitrided pistens were used. Previous runs.vith either stainless steel or~ chrome-plated stainless _ steel pistons have only averaged nine months.
' J.
A fuel channel that prevented locking of the grid bars durinE th'e sixth refueling outage was inspected. A small dent'(upproximately two inches long by one-half inch deep) was found on the bottom of the sup-port tube which fits the mushroom on the vessel. A means of repair is being considered and will be performed at a later date.
K.
During plant start-up on October 19, the poison system Valve CV-4050 (poison system inlet to RCP suction) did not open when the re-t_
actor recirculating pumps were started. Plant start-up was delayed until the actuator for this valve was repaired. Investigation revealed that the "0"-rings on the air. actuator (solenoid Valve SV h900) had 4
hardened, thus making the actuator inoperative. Please note that another piping route is available for the flow of poison to the reactor vessel a
if this line and/or the reactor recirculation pumps are not operable.
i V.
CHANGES. TESTS AND EXPERIMENTS PERFORMED PURSUAVf TO 10 CFR 50.59(a)
This section describer the changes made to the facility within
.the six-month period without prior Commission approval pursuunt to Section 50 59(a).of Title 10, Code of Federal Regulations to the extent that such changes constitute changes in the facility, as described in the Final Hazards Summary Report (FHSR). It also included tests and experiments carried out at the Plant without prior Commission approval pursuant to Section 50.59(a).
Each ' change, test or' experiment.is described as authorized only I
after a finding by Consumers Power Company that it did not involve a change
-1, Iin the Technical Specifications incorporated in Operating License DPR-6 l'
(effective May 1, 1964) or an unreviewed safety question.
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A.
Facility Changes 1.
A frequency meter was installed on the emergency diesel generator as an aid to the operators in monitoring the generator op-eration.
2.
A minor modification was added to the static power in-verter feeding Channel No 3 picoammeter. A voltmeter was added to the inverter front panel to allow monitoring of the output voltage.
Pre-viously, the voltage was monitored using a temporary meter.
3.
A shutoff valve for the steam drum conductivity cell was installed. This valve vill shut off the sample supply to the cell if a leak devel ps, such as sight glass breakage.
4 A start-up vant valve located on the shell side and adja-cent to the high-pressure feed-water heater was removed and replaced with a straight piece of piping. Steam leaks have constantly developed in this valve. The valve has always operated in the open position and has never been used for any operation in the closed position.
B.
Tests 1.
A water-quality test program is progressing at this writing and is being conducted in a joint effort between General Electric Company and Consumers Power Company. Test probes were installed in the con-densate feed-water system at the following locations:
a.
Feed-water header after the high-pressure feed-water heater.
b.
High-pressure feed-water heater shall drain header.
c.
Intermediate-pressure feed-water heater shell drain, d.
Inlet to the condensate demineralizers after the gland seal condensers.
e.
Outlet of the condensate demineralizer ahead of the low-pressure feed-water heater.
2.
The station power automatic transfer from the 138 kV line source to the h6 kV line source was tested during the sixth refueling outage. The No 2 condenser circulating pump was slow in returning to service (approximately 30 seconds instead of approximately 3.7 seconds).
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The ACB trip delay timer for this unit was tested and reset for about k.h seconds. All other automatic transfer equipment in service at that time worked properly.
3.
A temperature coefficient test was conducted in May prior to power operation with the newly load-d core. Test data indicated that the coefficient turned negative at 145 F after adding 14.25 cents of reactivity.
VI.
TRAINING Six hours of lecture and demonstrations were presented to each shift of the emergency room staff of both the Charlevoix and Little Traverse Hospitals. The lectures and demonstrations included Radiation Fundamentals, External Exposure and Limits, Internal Expocure and Contamination, Units and Definitions, Biological Effects, Background Radiation, Plant Organization and Radiation Protection Procedures, Instrument Demonstration and the Hos-pital Assistance Plan. The Hospital Assistance Plan was demonstrated by a test run utilizing a person contaminated specifically fcr this demonstration.
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10 VII. PERIODIC TESTING PEEFORMED AS REQUIRED BY THE TECHNICAL SPECIFICATIONS The following tabulation shows the required frequency of testing, plus the testing date of the systems or functions, which may be periodically tested per Technical Specifications:
System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives Continuous withdrawal and insertion Each major refueling and 10-18-69 of each drive over its stroke with at least once every six normal hydraulic system pressure.
months during periods cf Minimum withdrawal time shall be power operation.
23 seconde.
Withdrawal of each drive, stopping Each major refueling and 10-18-69 at each locking position to check at least once every six latching and unlatching operations months during periods of and the functioning of the position power operation.
indication system.
Scram of each drive 'from the fully Each major refueling and 10-18-69 withdrawn position. Maximum scram at least once every six 2
time from system trip to 90% of months during periods of insertion shall not exceed 2 5 power operation.
seconds.
Insertion of each drive over its Each major refueling but 5-1-69 entire stroke with reduced hydrau-not less than once a lic system pressure to determine year.
that drive friction is normal.
Control Rod Interlocks Rod withdrawal blocked when any Each major refueling but 4-30-69 two accumulators are at a pres-not less frequently than sure below 700 psig.
once every twelve months.
Rod withdrawal blocked when two Each major refueling but h-30-69 of three power range chonnels read not less frequently than below 5% on 0 - 125% scales (or once every twelve months.
below 2% on their 0 - h0% scales) when reactor power is above the minimum operating range of these channels.
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System or Function Frequency of Dates Undergoing Test
' Routine Tests Tested Control Rod Interlocks (Contd)
Rod withdrawal blocked when scram Each major refueling but 4-30-69.
dump tank is bypassed.
not-less than once every
. twelve months.
Rod withdrawal blocked when mode Each major refueling but 5-1-69 selector switch is in shutdown not less frequently than 4
position.
once every-twelve months.
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Other Liquid poison system component Two months or less.
7--1-69 check.
8-29-69 10-27-69 Post-incident spray system auto-At each major refueling h-29-69 matic control operation.
shutdown but not less frequently than once a year.
Core spray system trip circuit.
Not less frequently than 5-1-69 once every twelve months.
Emergency con' denser trip circuits.
Not less frequently than 5-8-69 once every twelve months.
Containment Containment sphere access air Once every six months or 10-10-69 locks and vent valves, leakage less.
rate.
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Isolation valve operability and At least once every h-20-69 leak tests.
twelve months.
-Isolation valve controls and Approximately quarterly.
7-9-69 instrumentation tests.
Penetration inspection.
At least once every h-29-69 twelve months.
Integrated leak test.
Once every two years.
7-12-69
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The following instrument checks and calibrations were performed
-at least once a month:
1.
Reactor safety system checks not requiring Plant shutdown.
2.
Air e,jector off-gas monitor.
3.
Stack-gas monitor calibration.
N.
Emergency condenser vent monitor.
5 Process monitor.
6.' Area monitoring system.
By
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Robert L. Haueter Electric Production Superintendent -
Nuclear.
Consumers Power Company Jackson, Michigan Date: -December 22, 1969 1
Sworn and subscribed to before me this 22nd day of December 1969.
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Notary Public, Jackson County Michigan My commission expires January 15, 1972
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