ML19351E228
| ML19351E228 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/31/1973 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8011260282 | |
| Download: ML19351E228 (8) | |
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f YANKEE NUCLEAR POWER STATION j
OPERATION REPORT #144 FOR TIIE MONTIl 0F DECEMBER l
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Submitted by YANKEE ATOMIC ELECTRIC COMPA!W Westboro, Massachusetts i
January, 1973 8011260 b b R
i This report covers the operation of the Yankee Atomic Electric Company at Rove, Massachusetts for the month of December 1972.
i The plant continued its maintenance outage throughout the report l
period.
Plant Abnormal Occurrences Abnormal Occurrence No. 72-20, Loss of Coolant on No. 3 Diesel Generator.
On December 27 at 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br /> during a surveillance test of No. 3 Emergency Diesel Generator the radiator hose became disconnected and coolant water was discharged to the floor.
The operator monitoring the test im-mediately shut down the diesel.
i After starting No. 1 and No. 2 Emergency Diesel Generators the radiator hose was reconnected and anti-freeze added to the No. 3 diesel.
The surveillance test was then successfully campleted on No. 3 Emergency Diesel Generator.
Plant Shutdowns l
Shutdown No. 132-10-72: October 18, 1972 continuing through the report period. A schedulad plant shutdown for replace-ment of control rods.
Inspections During the current naintenance outay the following inspections were made.
1.
The bolts joining the core barrel to the lower core support barrel vere all in place. No separation between the flanges was observed.
i 2.
The bolts joining the lover core support plate to the core barrel were all in place with no observable separation at the joint.
3.
The lower core support plate was measured for sag with a straightedge and fo';.nd to be level.
h.
The bolts joining the top of the shroud tubes to the bottom of the lover core support plate were inspected. Three bolts along with their locking cups were found missing from shroud 26 and four bolts as well i
as two locking cups were missing from shroud 27 The locking cup at position 8BC shoved some damage but both bolt and cup vere in place.
S.
All of the bolts holding the shroud tie plate to the individual shroud 4
tubes were inspected.
They appeared tight except for shroud 26; one of its two bolts was missing and the other had backed out part way.
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- i 6.
The shroud tio plate was intact except for one lateral bar between shrouds 2h and 25 which was cut with a gap of approximately one inch.
T.
The internal vertical dimensions of the eight shim rod shrouds and No. 17 and No. 18 control rod shrouds were measured and found to be close to the original design values.
8.
The thermal shield bolting, seam clamp assemblies and secondary core support assemblies were normal.
9 The cladding defect on the bottom of the reactor vessel was inspected with television and binoculars.
No pitting of the base metal or undercutting of the cladding was discernable.
10.
The eight shim rods were inspected. The vane surfaces, edges and bottom were inspected by television. The overall length of each shim rod was measured. The au found length differed from design values up to 0.40 inches ( 0.062 inches). The overall bow of the shim rods as determined with a transi-varied from negligible to approximately 0.5 inch.
11.
The overall bow of eight control rods as determined with a transit varied from negligible to approximately 0.5 inch.
12.
Eleven ruel assemblies were inspected with the underwater periscope.
No unusual conditions were found. The fuel assembly (Ah21U) that had i
been damaged during removal was inspected by the manufacturer in the spent fuel pit.
The damage was sufficient to preclude return of the assembly to the core.
13.
An outside contractor gamma scanned the four fuel pins previously removed from assembly BL36U by the manufacturer.
No abnormalities were seen.
Preliminary analysis indicates approximately 1% fuel densi-fication in the pins.
Plant Maintenance The following is a list of pertinent plant maintenance items performed by the plant staff during the month of December, 1972.
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1.
The gravity drain tank pumps were repaired.
2.
Repairs to No. h main coolant check valve were completed.
Instrumentation and Control The following is a list of pertinent instrumentation and control maintenance items performed by the plant staff during the month of 4
December, 1972.
1.
The loop seal radiation monitor detectors were replaced.
N The following primary channels were recalibrated:
2.
No. 3 and No, h main coolant loop pressure, a.
b.
Pressurizer pressure.
c.
No. 1, No. 2, Nc. 3 and No. k main coolant flow.
3.
The safety injection status indicators were modified to reduce heat buildup and to increase the lamp life.
Chemistry All fuel was removed from the reactor core and stored in the Spent Fuel Pit during the previous report period. Consequently chemistry parameters for the main coolann have been omitted from this report.
Health and Safety Waste disposal liquid releases totalled 51,131 gallons containing 0.025 me of gross beta-Samma activity and 17 0h curies of tritium.
In addition 1.68 curies of tritium was released as a vapor. Secondary plant water discharged totalled 27,000 gallons containing 0 791 curies of tritium.
Radiation exposure doses for Yankee personnel, NEPSCo personnel
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and Outside Contractor personnel as measured by film badge for the month of Decenber,1972 were as follows :
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Yankee Plant Personnel Average accumulated exposure dose:
Sh mrem Maximum accumulated exposure dose:
h80 mrem NEPSCo Personnel Average accumulated exposure dose:
6 mrem Maximum accumulated exposure dose:
20 mrem Outside Contractor Personnel Average accumulated exposure dose:
265 mrem Maximum accumulated exposure dose: 2,7h0 mrem Operations Attached is a su= mary of plant operating statistics and a plot of daily average load for the month of December, 1972.
The following Plant Design Change Requests completed in 1972 were reported in Operations Report No 143 for November, 1972. They are reported again to incorporate a Safety Analysis for each.
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72-1 Diesel Oil Makeup - Controls Power Supply. This change concerned providing an' additional 480/120 volts ac transformer connected to No. 2 Auto-Throvover switch to provide power for the diesel oil makeup controls. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR based upon the following:
With an extended total loss of ac the emergency diesels will run out of fuel if the makeup controls are not supplied from an automatic throwover supply.
72-3 Change of Pressurizer Wide Range Level Channel from No. 3 to No. 2 Primary Instrument Power Supply.
This change involved connecting the wide range pressurizer level channel to a separate power supply from the narrow range channel. The safety evaluation concluded that the change did not present any significant hazard not described l
or implied in the FHSR based upon the following:
Because of the importance of having at least one means of 0
pressurizer level indication at all times, it is deemed necessary to provide as much separation between the two channels as possible. Eliminating the use of a common current reference will completely separate their power supplies.
4 72 h Vital Eus Cabinet Wiring. This change involved separation of the motor-generator wiring from station service wiring within the vital bus control cabinet. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR baced upon the following-Eecause of the random interlacing of the energized wires within the vital bus cabinet it will be difficult and hazardous to make any repairs to the control circuits.
With the change vital instrumentation loads can be maintained I
while the vital bus control circuits are being repaired.
72-7 Control Board Ventilation Change. This change involved blocking the openings leading from the main control board to the switchgear roca ventilation system.
An air conditioner was installed to compensate for the lack of ventilation. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR based upon the following:
3 The interior of the main control board has many openings leading to the switchgear room to provide ventilation.
In the event of a fire in the switchgear room the smoke and heat will pass up through the interior of the main control board to the false ceiling area of the control room. With the addition of the change, the smoke and heat will be forced out into an open area in the turbine hall.
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-9 72-8 Check Valve Addition to Vapor Container Recirculation System Piping.
This change involved adding a check valve in the V.C. recirculation system to prevent over pressurizing the purification pumps discharge header during monthly recirculation valve operational checks. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR based upon the following:
During normal operation the LPSI accum -
or will pressurize the low pressure safety injection piping ;o 420 psig. Monthly V.C. Recirculation motor operated valve exercises vill result in pressurizing the purification pump discharge header 2 ",
schedule 10 piping to this pressure. The addition of the check valve vill prevent this over pressurization.
72-9 Main Steam Line Safety Valve. This change involved the additions of nozzle supports to the main steam line safety valves to increase the margin of safety during relief operation. No safety evaluation was required for this change.
72-16 Low Main Coolant Flow Trip System Modifications. This change involved the correction of a wiring deficiency which prevented the turbine trip signal, caused by a 1 css of main coolant flow, from reaching the turbine trip valve. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FESR based upo., the following:
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The changes to enable a 1cv main coolant flow condition to trip the turbine-generator via the permissive and scram auxiliary relays does not present an unreviewed safety analysis.
72-17 Safety Injection Accumulator Hitrogen Regulator Pilot Replacement.
This change involved replacing the two nitrogen regulator pilot valves rated at 1200 psig with valves rated at 1500 psig.
The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR based upon the following:
Use of the Model 825 pilot valvts rated at 1500 psig in place of the Model 820 pilot valves rated at 1200 psig vill not adversely affect the proper and timely operation of the full ficw nitrogen regulators of the low pressure safety injection accumulator.
72-18 Redundant S. I. Accumulator Tank Level Indication. The change in-volved the installation ef' a second S.I. accumulator level transmitter and indicator to ensure
- u. accurate knowledge of the level at all times. No safety evaluation was required for this change.
72-19 Main Coolant Flow Transmitter Vent Modifications. This change was 4
replaced by Engineering Design Change Request No. 72-7 described below.
72-30 Charging Line Replacement at Vapor Container Penetration. This change involved replacement of approximately eight feet of 2 inch charging line pipe and the vapor container penetration cap.
The safety evaluation concluded that the change did not present any significant hazard not described in the FHSR based upon the following:
This repair will not affect the proper operation of the charging system. This portion of the charging system is a Safety Class 2 system.
Engineering Dasign Change Request No. 72-T, Main Coolant Flow Transmitter Vent Modification was completed. This EDCR replaced PDCR No. 72-19 above.
This change involved replacement of the operating handles on the vent valves with a modified design. The safety evaluation concluded that the change did not present any significant hazard not described or implied in the FHSR based upon the following:
The replacement of the tee operating handles for the vent valves and the installation of threaded inserts vill de-crease the potential for leakage of the main coolant flow transmitter vents. Retapping the high pressure vent valve threads will not reduce tre body thickness below the re-quired value.
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YANKEE ATOMIC ELECTRIC COMPANY - OPERATING SU:C'ARY DECEMBER 1972 ELECTRICE MONTH YEAR TO DATE Gross Generaticn KWH 0
690,ll3. i'i o 13,705,989,100 Sta. Service (While Gen. Incl. Losses)
KWH 0
45,629,191 886,698,805 Net Output KWH 0
6hh,h84,009 12,819,290,295 Station Service 0
6.61 6.h7 Sta. Service (While Not Gen. Incl. Losses) KWH 505,156 6,765,51h 39,869,906 Ave. Gen. For Month (7hh)
KW 0
Ave Gen. Running (0)
KW 0
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PLANT PEi ')RMANCE Net Plant Efficiency 0
28.80 28.50 Net Plant Heat Rate btu /KWH N/A ll,8h9.83 11,974.56 Plarrt Capacity Factor O
42.45 7h.00 0
Reactor Plant Availability 0
55 07 82.81 NUCLEAR MONTH CORE X TOTAL Hours Critical HRS 0
3,75h.15 89,239..h Times Scrammed 0
9 70 4
Burnup Core Average MWD /LifU 0
3,726.81 Region Average MWD /MfU l
A (INNER) 0 3,279 05 21,634.55 B (MIDDLE)
O h,080.64 14,051'.99 C (0 UTER-ZIRCALOY) 0 3,380.43 3,380.h3 i
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YANKEE ATOMIC EI+mIC COMPANY 6
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