ML20099E474
| ML20099E474 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/31/1984 |
| From: | Ronafalvy J, Zupko J Public Service Enterprise Group |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NUDOCS 8411210325 | |
| Download: ML20099E474 (20) | |
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L- AVERAGE DAILY UNIT POWER LEVEL Docket No. 50-272 Unit Name Salem # 1
.Date Nov. 10,1984.-
- Completed by -J. P. Ronafalvy Telephone 609-935-6000 Extension 4455 Month October 1984 Day Average Daily Power Level Day' Average Daily Power Level (MWe-NET) (MWe-NET) 1- O. 17 0 2- 0 18 0 3- 0 19 0 4 0 20 0 5- '0 '21 0 6 0 22 23 7 0 23 0 8 0 24 304 9 0 25 337 10 0 26 465 f
11 0 27 626 12 0 28 678 13 0 29 836 i 14 0 30 311 15 0 31 352 16 0 P. 8,1-7 R1 h
b 8411210325 841031 [ '/
PDR ADOCK 05000272 R pop ,
Page 2 of 20
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OPERATING DATA REPORT
, Docket No._50-272 Date Nov. 10, 1984 ~
Telephone 935-6000 Completed by J. P. Ronafalvy Extension 4455 Operating Status 1, Unit Name Salem No. 1 Notes
- 2. Reporting Period October 1984
- 3. Licensed Thermal Power (MWt) 3338
- 4. Nameplate Rating (Gross MWe) 1135
- 5. Design Electrical Rating (Net MWe) 1090
- 6. Maximum Dependable Capacity (Gross MWe)1124
- 7. Maximum Dependable Capacity (Net MWe) T679
- 8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give Reason N/A
- 9. Power Level to Which Restricted, if any (Net MWe) N/A
- 10. Reasons for Restrictions, if any N/A This Month Year to Date Cumulative
- 11. Hours in Reporting Period 745 7320 64345
- 12. No. of Hrs. Reactor was Critical 368.8 1606.4 34757.6
- 13. Reactor Reserve Shutdown Hrs. 0 54.5 3088.4
- 14. Hours Generator On-Line 205.6 1403.4 33181.3
- 15. Unit Reserve Shutdown Hours 0 0 0
- 16. Gross Thermal Energy Generated
(!001) 389047 4189070 100008441
- 17. Gross Elec. Energy Generated (bOGI) 104040 1385420 33000520
- 18. Net Elec. Energy Generated (MWH) 77715 1268001 31239313
- 19. Unit Service Factor 27.6 19.2 51.6
- 20. Unit Availability Factor 27.6 19.2 51.6
- 21. Unit Capacity Factor (using MDC Net) 9.7 16.1 45.0
- 22. Unit Capacity Factor (using DER Net) 9.6 15.9 44.5
- 23. Unit Forced Outage Rate 72.4 71.8 33.3
- 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
N/A
- 25. If shutdown at end of Report Period, Estimated Date of Startup:
- 26. Units in Test Status (Prior to Commercial Operation) :
Forecast Achieved Initial Criticality 9/30/76 12/11/76 Initial Electricity 11/1/76 12/25/76 Commercial Operation 12/20/76 6/30/77 8-1-7.R2 Page of Page 3 of 20
UNIT SHUTDOWN AND POWER REDUCTIONS Dock 3t No',50-272 REPORT MONTH October 1984 Unit N;me Salem No.1' Date Nov. 10,1984 -
Completed by J.P. Ronafalvy Telephone 609-935-6000 -
Extension 4455 Method of Duration Shutting License Cause and Corrective No. Date Type Hours Feason Down Event System Component Action to 1 2 Reactor Report Code 4 Code 5 Prevent Recurrence Nuclear Other Control 84-180 9-10 F 207 A 4 -
RB CRDRVE Rod Drive Problem Nuclear Core Physics84-182 10-9 S 200.6 B 4 -
RC ZZZZZZ Test Seal Oil System and 84-184 10-17 F 100.2 A 4 -
HA XXXXXX Seals Generator Turbine Overspeed 84-186 10-22 F 1.2 B 4 -
HA ZZZZZZ Trip Test 84-188 10-22 F 30.4 A 3 -
HA INSTRU Turbine Instruments Nuclear Core 84-190 10-24 S 32.8 B 5 -
RC ZZZZZZ Physics Test Steam Generator Feed 84-192 10-25 F 22.7 A 5 -
HE PUMPXX Pump Problems Condensate /
84-194 10-28 F 5.9 A 5 -
HH PUMPXX Hotwell Pumps Loss of Vacuum /High l 84-196 10-28 F 51.1 A 5 -
HA XXXXXX Back Pressure Reactor Coolant Pump 84-200 10-30 F 7.2 A 5 -
CB INSTRU Instrumentation Loss of Vacuum /High 84-202 10-30 F 27.0 A 5 -
HA XXXXXX Back Pressure 1 2 Reason 3 Method 4 Exhibit G S Exhibit 1 F: Forced A-Equipment Failure-explain 1-Manual Instructions Salem as y S: Scheduled B-Maintenance or Test 2-Manual Scram. for Prepara- Source D C-Refueling 3-Automatic Scram, tion of Data
$ D-Regulatory Restriction 4-Continuation of Entry Sheets u E-Operator Training & Licensing Exam Previous Outage for Licensee F-Administrative 5-Load Reduction Event Report G-Operational Error-explain 9-Other. (LER) File g H-Other-explain (N'U REG 0161) o
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> ~ MAJOR PLANT MODIFICATIONS DOCKET-NO.: 50-272
, REPORT MONTH October 1984 UNIT NAME: Salem 1 DATE: November 10,'1984 COMPLETED BY: J. Ronatalvy TELEPHONE: 609/339-4455
- DCR NO . PRINCIPLE SYSTEM . SUBJECT 1ET-1166 Incore Instrumentation Disconnect the.P-250 computer from-the incore flux mapping.
system.
1EC-1231 Waste Disposal Liquid Add break flanges to the casing drain lines of #11 and
- 12 Reactor Coolant Drain Tank pump casing.
1EC-1641 Containment Ventilaton Install a 10" 'T' connection between IVCS and the 10" Auxiliary Building ducting.
1EC-1665 Component Cooling Retube No. 11 Component Cooling Heat Exchanger with titanium tubec (or suitable available material) .
lEC-1728 Chemical and Volume Replace the piping downstream-Control of the 1CV45 and ICV 50 valves, located on #11 and $12 C/SI pump's. casing drain lines, with piping IAW Pipe Spec 496.
Install 1500# blind flanges at the end of these' lines.
Fabricate spool pieces to connect the casing drain lines to their respective floor drain lines. The spool pieces are to be used during maintenance on the pumps and are to be in accordance with Spec.
1EC-1735 Security System Install high mast (100')
lighting fixtures and upgrade existing yard and perimeter lighting.
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_p- 4 MAJOR' PLANT-MODIFICATIONS -DOCKET NO.: 50-272
- REPORT MONTH October'1984 UNIT NAME: -Salem-1
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,DATE: : November 10,;1984
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COMPLETED BY:~ J. Ronatalvy TELEPHONE:' 609/339-4455
- DCR NO.. PRINCIPLE. SYSTEM SUBJECT 1EC-1745 Main Generator-' Add additional [ piping'in the;
' Hydrogen
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main; generator and between.the generator and hydrogen dryer to provide adequate differential pressure ^across the hydrogen. dryer.
1EX-1759 SEC (R130) . Provide temporary instrumentation to various inputs.to the lc SEC.
1EC-1809 Reactor Coolant-Pump.- Modify the shaft vibration Shaft Vibration monitor system.to eliminate Monitors erroneous vibration monitor readings.
1EC-1817 Diesel Generator E300 -Replace existing transformer-T54 in.the exciter regulator cubicle with a:new type.:
1EC-1849 Main Steam Install additional steam stop valve hydraulic control bypass valve trouble light on control console for each of four-valves'11-14MS167.
1ET-1856 . Main Generator Loop test.
1EC-1862 Safety Injection Replace existing lube oil coolers on 11 and 12 SI Pumps with coolers'of similar design characteristics but upgraded materials of constructoin.
Upgraded coolers will have outer tubesheets and tubes consisting of titanium material.
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' MAJOR PLANT MODIFICATIONS DOCKET NO.: 50-272 REPORT MONTH October 1984' UNIT NAME: Salem-1 DATE: November 10, 1984 COMPLETED BY: J. Ronafalvy TELEPHONE: 609/339-4455
- DCR NO. PRINCIPLE SYSTEM SUBJECT i.
1EC-1870 Radiation Monitoring Provide an~ area radiation monitor in the Electrical.
Penetration Area with indication in the C with a range of 10 gntrol to 10goom R/HR.
1EC-1907 Safety Injection- Replace existing SI' throttling valves - (i.e. , 11-14SJ16) with needle valves which are designed for metering flow.
An additional globe valve will be installed.in series with each SJ16, and will be used as an isolation valve during shutdown.
lEC-1956 Pressurizer System Modify Control Circuit to IPR 6 and 1PR7 valves.
1SC-1167 Reactor Coolant Change reactor coolant flow transmitter from Fisher Porter model #10B2496PB to new model and type specified by Engineering.
1SC-1331 Chilled Water Install an additional suction gauge no greater than 40 PSI for 11 and 12 Chilled Water Pumps.
1SC-1378 #1 Main Generator Install imbedded slot temperature detectors in #1 Main Generator.
Page 7 of 20
4 MAJOR PLANT MODIFICATIONS DOCKET NO.: ~ 50-272
, REPORT-MONTH OCTOBER 1984 UNIT NAME: Salem 1 DATE: November 10, 1984 COMPLETED BY: J. Ronafalvy TELEPHONE: 609/339-4455
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- DCR NO. SAFETY EVALUATION . 10 CFR 50.59 1ET-1166 -This change permits'a portable computer to obtain data
.from non-safety grade instrumentation. The instrumentation is used_to obtain periodic test data (flux map of the core) and is not involved in process control.
No'unreviewed safety or environmental questions are involved.
1EC-1231. This. change does r.ot alter _ the original design concept of the piping system in any way for the_ Reactor Coolant Drain Tank Pump. Also, this change does not alter the Technical Specification or the FSAR and will not increase the liquid effluent discharge.from the Station. No unreviewed safety or environmental questions are involved.
1EC-1641 This design modification will not affect the ability of valve IVC 5 to perform its Containment isolation >
requirement. Also, the design modification does not create the possibility for an accident or malfunction of a different type than any previously evaluated and does not ;
reduce the margin of safety defined in the basis of any Technical, Specification. No unreviewed safety or environmental questions are involved.
1EC-1665 This design change involves a change of materials to an upgraded type. This modification will not alter any plant process or discharge and will not affect the existing plant impact. No unreviewed safety or environmental questions are involved.
1EC-1728 .This change only involves addition of a revised piping diagram. No unreviewed safety or environmental questions are involved.
IEC-1735 This change only involves an upgrade of the yard lighting system. No unreviewed safety or environmental questions are involved.
- DCR - Design Change Request 1
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- DCR NO. SAFETY EVALUATION 10 CFR 50.59 lEC-1745 This design change involves the addition of small bore piping in the Main Generator Hydrogen System to facilitate adequate hydrogen flow. It does not affect Plant operating procedures. No change to the FSAR or the Technical Specifications is required. No unreviewed safety or environmental questions are involved.
LEX-1759 The implementation of this DCR requires the addition of a small amount of fire retardent wood. The area is fire protected. The amount of wood is below the design basis for fire protection. No unreviewed safety or environmental questions are involved.
1EC-1809 The implementation of this DCR involves penetration of a fire barrier. Instructions are included for proper resealing to maintain the required hourly fire rating. No unreviewed safety or environmentel questions are involved.
1EC-1817 The replacement component meets or exceeds the design of the old part. Qualification includes requirements of IEEE 344 for seismic events. No unreviewed safety or environmental questions are involved.
1EC-1849 the intended function of the system remains unchanged. No unreviewed safety or environmental questions are involved.
1ET-1856 This DCR is part of the Westinghouse inspection, overhaul and rewind of the Unit 1 Main Generator. No unreviewed safety or environmental questions are involved.
1EC-1862 This DCR documents the replacement of the SI Pump Lube Oil Coolers with upgraded materials. The new coolers are seismically qualified to seismic I criteria and the ASME Section III Class 3 - 1981 Summer Addends for the tube side only. The shell side is classified to the, 1980 ASME Section VIII Code Division I. This DCR docs not increase the potential for an accident nor does it degrade the integrity of the Service Water System and the Lube oil System for the bearings on the SI Pumps. No unreviewed safety or environmental questions are involved.
- DCR - Design Change Request Page 9 of 20 J
- DCR NO. SAFETY EVALUATION 10 CFR 50.59 1EC-1870 This DCR adds a radiation monitor to determine the magnitude of a release of radioactive material in the electrical penetration area. Information from this monitor is not required for the safe shutdown of the Unit.
No unreviewed safety or environmental questions are involved.
1EC-1907 This design change replaces the boron injection path throttle valves with ones of a more applicable design.
Also, additional valves for isolation are included in this DCR. This design does not alter the intent of the Safety Injection System. It does not increase the consequences of an event, nor the likelihood of an occurrence. No unroviewed safety or environmental questions are involved.
1EC-1956 This design change involves rewiring for enhanced valve operation as recommended by the manufacturer. No unreviewed safety or environmental questions are involved.
1SC-1167 This DCR involves replacement of an existing transmitter.
No unroviewed safety or environmental questions are involved.
1SC-1331 This DCR has installed a new gauge as por ASME Section XI.
No unreviewed safety or environmental questions are involved.
ISC- 13 7 S This DCR describes the addition of forty-eight (48) thermocouples for trending information of the generator.
No unreviewed safety or environmental questions are involved.
- DCR - Design Change Request f
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PSE&G SALEM GENERATING STATION -
SAFETY RELATED WORK ORDER LOG SALEM UNIT I WO NO DEPT UNIT EQUIPMENT IDENTIFICATION 949975 MD 1 #12 CHARGING PUMP FAILURE DESCRIPTION: INSPECT CASING ON PUMP; EVALUATE CLADDING CRACK.
CORRECTIVE ACTION: REPLACED PUMP WITH NEW PUMP.
I 84-09-27-103-6 SMD 1 #15 CFCU FAILURE DESCRIPTION: SERVICE WATER LEAK ON MOTOR COOLER LINE.
CORRECTIVE ACTION: WELDED NEW FLANGE, PIPE, AND ELBOW; INSTALLED NEW PIPING AND GASKETS l
84-09-21-038-0 l
SMD 1 #11 CFCU SERVICE WATER FAILURE DESCRIPTION: 3/4" HEADER HAS LEAK ON THE COIL SIDE OF 11SW248 (VENT VALVE)
CORRECTIVE ACTION: GROUND OUT FLANGE AND PIPE; REWELDED NEW PIPE AND l
FLANGE- INSTALLED NEW GASKET 0099129183 SMD 1 II1 CFCU FAILURE DESCRIPTION: SW LEAK FROM TELLTALE ON MOTOR COOLER (WATER
! m SPRAYING WITH UNIT IN SERVICE)
!d 0 CORRECTIVE ACTION: INSTALLED NEW COOLING COIL P*n M
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SALEM UNIT 1 ;
WO NO DEPT UNIT EQUIPMENT IDENTIFICATION 84-09-27-114-1 SMD 1 #12 CFCU FAILURE DESCRIPTION: PIPE UPSTREAM OF THE 12SW405 AND THE FLOW TAPS ON THE OUTLET OF THE CFCU IS LEAKING. i CORRECTIVE ACTION: GROUND OUT HOLE IN LINE TO SOUND METAL; PAD WELDED AS PER ENGINEERING 84-09-20-028-1 SIC 1 BIT RECIRCULATION FLOW INST.
FAILURE DESCRIPTION: NO OHA ON LOW FLOW; FACEPLATE IS BROKEN (BIT INOPERABLE)
CORRECTIVE ACTION: REPLACED FLOAT BODY ASSEMBLY, FLOAT GUIDE, RETAINING RING FLOAT, EETAINING RING TUBE,.AND FLEXITALLIC GASKET 0099104776 SMD 1 BIT RECIRCULATION LINE FAILURE DESCRIPTION: THE LINE IS CLOGGED BETWEEN VALVES ISJ79 AND CVl61 CORRECTIVE ACTION: CUT LINES, CLEARED BLOCKAGE, AND REWELDED LINE 943625 NCS 1 VALVE ISAll4 FAILURE DESCRIPTION: VALVE FAILED LEAK RATE TEST
, CORRECTIVE ACTION: LAPPED AND BLUE CHECKED VALVE AND REPLACED BONNET
, y GASKET o
y 943684 NCS 1 VALVE 11SS908 FAILURE DESCRIPTION: VALVE LEAKED DURING TESTING N CORRECTIVE ACTION: RENEWED PACKING
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SALEM UNIT 1 ' .:
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DEPT UNIT EQUIPMENT IDENTIFICATION y WO NO 84-06-07-990-8 NCS 1 VALVE 13SS182 1
l i FAILURE DESCRIPTION: V.'.LVE FAILED LEAK RAT" TEST CORRECTIVE ACTION: REPLACED PACKING; BLUE CHECKED, LAPPED SEAT, AND LAPPED PLUG l
943760 NCS 1 VALVE INT 26 FAILURE DESCRIPTION: VALVE FAILED LEAK RATE TEST CORRECTIVE ACTION: REPLACED BONNET GASKET; LAPPED AND BLUE CHECKED 940829 MD 1 #12 AUXILIARY BUILDING SUPPLY FAN i
FAILURE DESCRIPTION: HEATING. COILS RUPTURED CORRECTIVE ACTION: CRIMPED TUBING OF COILS AND' SILVER BRAZED;-TESTED' WITH AIR AND SNOOPED 943631 10 1 VALVE #12GB3 FAILURE DESCRIPTION: VALVE FAILED LEAK RATE TEST l
CORRECTIVE ACTION: VALVE GASKET SURFACE MACHINED 0099100339 SMD 1 #11 CHILL WATER PUMP FAILURE DESCRIPTION: TRIPS ON OVERLOAD 5 INSTALLED NEW MOTOR; ECM LINED MOTOR TO PUMP O CORRECTIVE ACTION:
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I SALEM' UNIT 1 -
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1 5 WO NO DEPT UNIT EQUIPMENT IDENTIFICATION I
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.84-10-14-009-8 SMD 1 1CCl31 LIMITORQUE 1 ~i i FAILURE DESCRIPTION: VALVE FAILED TO MAKE OPEN LIMIT AND A BURNING: SMELL i WAS NOTED IN THE VICINITY OF THE BREAKER
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CORRECTIVE ACTION: REPLACED MOTOR ON LIMITORQUE; FIXED BROKEN' ARM'ON BREAK ASSEMBLY ., .
L 009900281-7 SMD- 1 #12 SERVICE WATER PUMP REPLACE AIR RELEASE VALVE AS PER DR:#MD.84-3313 I
FAILURE DESCRIPTION:
CORRECTIVE ACTION: REPLACED VALVE AND TIGHTENED FITTINGS' '
.I 84-09-29-007-3 j- SMD 1 #12 SERVICE WATER PUMP FAILURE DESCRIPTION: SEVERE PUMP PACKING. LEAK CORRECTIVE ACTION: REPACKED PUMP AND STRAINER WITH BALZONA METAL KIT f ON AIR RELEASE VALVE 1 84-09-12-004-6 l SMD 1 #13 SERVICE WATER PUMP
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j FAILURE DESCRIPTION: UPPER MOTOR BEARING OIL LEAK k
i CORRECTIVE ACTION: REPLACED MCYTOR E 84-08-04-325-1
{ $ SMD 1 VALVE filSW312 (SCREEN WASH DRAIN' VALVE)
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5 FAILURE DESCRIPTION: ' VALVE IS ERODED AND LEAKS .
l- O WELDED IN NEW PIPE AND REPLACED VALVE-1 " CORRECTIVE ACTION:
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-q SALEM UNIT 1 ,.
WO NO DLYI UNIT EQUIPMENT IDENTIFICATION 930189 NCS 1 VALVE ISW26 FAILURE DSSCRIPTION: VALVE LEAKS CORRECTIVE ACTION: REPLACED VALVE WITH NEW 30" PRATT BUTTERFLY VALVE 922465 NCS 1 VALVE 13SW20 FAILURE DESCRIPTION: VALVE LEAKS CORRECTIVE ACTION: REPAIRED RUBBER LINING AS PER ENGINEERING-INSTRUCTIONS-0099128853 SMD 1 #11SW23 CONTROL ROOM INDICATOR FAILURE DESCRIPTION: LIMIT-SWITCH STICKS WHEN OPENING VALVE CORRECTIVE ACTION: CLEANED CONTACTS AND REPLACED 33Y-3 RELAY l
l 009910958 SIC 1 VALVE 11MS10 FAILURE DESCRIPTION: VALVE WON'T STAY IN AUTO i
CORRECTIVE ACTION: REPLACED AUTO / MAN MODULE; CLEANED GEAR SHAFTS 0099128748 SIC 1 VALVE 11MS10 FAILURE DESCRIPTION: VALVE WON'T OPEN
. tl CLEANED PLUGGED ORIFICE IN E/P Ij CORRECTIVE ACTION:
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- s. ~L SALEM UNIT 1 -
EQUIPMENT IDENTIFICATION WO NO DEPT UNIT
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84-10-17-004-3 SMD 1 PRESSURIZER HEATERS FAILURE DESCRIPTION: DEIONS 4, 5, 6 ON BACKUP GROUP 11 (lGP3X) ARE '
TRIPPING OFF CORRECTIVE ACTION: REPLACED BREAKER 009900261-2
. SMD 1 INNER CONTAINMENT DOOR EL. 100' FAILURE DESCRIPTION: REQUIRES REPAIR CORRECTIVE ACTION: FLANGE BLOCK BEARINGS TIGHTENED
! 84-08-10-649-0 .)
SMD 1 100' EL. AIRLOCK i
FAILURE DESCRIPTION: OUTER SEAL DID NOT SEAL DURING CHECK CORRECTIVE ACTION: REPLACED WITH NEW SEALS AFTER CLEANING SEAL AREA 0099100771 SMD 1 #1 ALT SHUTDOWN SYSTEM INVERTER
- FAILURE DESCRIPTION
- INVERTER DOES NOT WORK CORRECTIVE ACTION: REPLACED C2 CAPACITOR; OUTPUT TESTED-l 0099122146 J
SMD 1 REACTOR TRIP BREIJER SYSTEM I
c fl FAILURE DESCRIPTION: BY-PASS BREAKER "A" FAILED TO CLOSE DURING P-4 D TESTING
- w f CORRECTIVE ACTION: INSTALLED NEW SWITCH ON CONSOLE #GE-10CP264SBY m
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9-SALEM UNIT 1 '?'
q WO NO DEPT UNIT EQUIPMENT IDENTIFICATION i 84-10-11-008-3 SMD 1 VALVE #13MS171 .
I FAILURE DESCRIPTON: STEAM LEAKING PAST SEAT .
CORRECTIVE ACTION: BLUE CHECKED; REPLACED GASKETS (VALVE WAS MISSING ~
SEAT RING GASKET) 84-05-10-216 SIC 1 VALVE 12MS169 FAILURE DESCRIPTION: VALVE DIAPHRAGM REQUIRES REPLACEMENT CORRECTIVE ACTION: INSTALLED NEW DIAPHRAGM'AND CHECKED VALVE STROKE AND SEATING 84-05-219-1 SIC 1 VALVE #14MS169 (STOP VENT VALVE)
FAILURE DESCRIPTION: VALVE DIAPHRAGM REQUIRES REPLACEMENT CORRECTIVE ACTION: REPLACED DIAPHRAGM AND STROKED VALVE 0099130131 SMD 1 PR3, 4, AND 5 FAILURE DESCRIPTION: OVERHEAD ALARM NOT FULLY SEATED (FLASHES)
CORRECTIVE ACTION: INSTALLED NEW CONTACT BLOCK AND CONTACT ARM; CHECKED CONNECTIONS; CHECKED TERMINATIONS; y
CALIBRATED o
$ 84-10-17-003-5 SMD 1 PRESSURIZER HEATERS o FAILURE DESCRIPTION: DEIONS 7, 8, 9'ON 11 BACKUP GROUP HEATERS (lGP3X) *
- ARE TRIPPING OFF
$ CORRECTIVE ACTION: REPLACED BREAKER
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l WO NO DEPT UNIT EQUIPMENT IDENTIFICATION d
84-09-10-001-1 SMD 1 1A DIESEL GENERATOR AUTO BEZEL ALARM FAILURE DESCRIPTION: ALARM CANNOT BE ACKNOWLEDGED WHEN DIESEL IS TAKEN TO MANUAL CONTROL CORRECTIVE ACTION: REPLACED-OPERATE / RESET COIL (STRUTHERS/DUNN 255XCX-P) 009910291 SMD 1 1C SEC FAILURE DESCRIPTION: TEST 18 WILL NOT. RESET CORRECTIVE ACTION: REPLACED CARD ASSEMBLY WITH SPARE;' REPLACED XK6 RELAY
- i 009902271 SIC 1 NIS CH N-35
-10 FAILURE DESCRIPTION: READING OF 10 AMP WITH NO FUEL IN CORE CORRECTIVE ACTION: -REPLACED DETECTOR-84-06-19-762-5 SIC 1 AUDIO COUNT RATE SCALER FAILURE DESCRIPTION: CHANNEL FAILED CORRECTIVE ACTION: -REPLACED TIMER / SCALER WITH OLDER MODEL ca 3 0099128P37 O SIC 1 SOURCE RANGE CHANNEL I N .
COUNTS WENT.UP BY;A FACTOR OF 10 WHILE CHANNEL II P FAILURE DESCRIPTION:
g -STAYED CONSTANT CORRECTIVE ACTION: REPLACED CONNECTOR ON CABLES; REPLACED PRE-AMP
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SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
- UNIT NO. 1 OCTOBER 1984 Unit No. 1 began the period shutdown as the fifth refueling outage drew to a close. The Unit entered Mode 3 on 10/05/84 at 2013 hours0.0233 days <br />0.559 hours <br />0.00333 weeks <br />7.659465e-4 months <br />.
Repairs to a leaking pressurizer spray valve and a faulty source range channel delayed heatup and subsequent testing. On 10/13/84 at 1635 hours0.0189 days <br />0.454 hours <br />0.0027 weeks <br />6.221175e-4 months <br /> the Unit entered Mode 2 (reactor critical) for Low Power Physics Testing. On 10/14/84 at 1605 hours0.0186 days <br />0.446 hours <br />0.00265 weeks <br />6.107025e-4 months <br /> the reactor was shutdown to Mode 3 in accordance with Technical Specifications because of an inoperable Containment Isolation Valve ICC131. Investigation revealed that the limitorque operator had failed and required replacement. On 10/14/84, in preparation for a reactor startup, testing revealed problems with one of the two manual reactor trip switches on the Control Room console which was subsequently' replaced. On 10/15/84 at 2244 hours0.026 days <br />0.623 hours <br />0.00371 weeks <br />8.53842e-4 months <br /> the Unit re-entered Fode 2 to continue Low Power Physics Testing which was completed on 10/16/84. On 10/16/84 No. 13 Condensate Pump lower motor bearing failed requiring replacement of the motor. As a result of problems with the Generator Seal Oil System during steady state roll at 1800 rpm, it was decided to inspect the generator.
Investigation revealed the most likely cause was a failure of the 12 psig regulator and a plugged cunofilter on the discharge side of the Seal Oil Pumps. At 0317 hours0.00367 days <br />0.0881 hours <br />5.241402e-4 weeks <br />1.206185e-4 months <br /> on 10/21/84 the reactor was brought subcritical (Mode 3) due to three of four DF13 valves closing without l cause. After extensive investigation, the valves retested satisfactorily and the reactor was brought critical on 10/21/84 at
- 2128 hours0.0246 days <br />0.591 hours <br />0.00352 weeks <br />8.09704e-4 months <br />. The Unit was synchronized at 0343 hours0.00397 days <br />0.0953 hours <br />5.671296e-4 weeks <br />1.305115e-4 months <br /> on 10/22/84 and ran for a minimum of eight (8) hours prior to a scheduled Turbine overspeed Test. At 1544, with the Unit load removed, the reactor tripped following completion of the Turbine overspeed Test. The trip was caused by induced vibrations in the First Stage Pressure Transmitter (PT506) resulting in a false pressure signal which armed a reactor protection circuit (P-7). Following replacement and calibration of PT506, the reactor was brought critical and the Unit synchronized at 2207 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.397635e-4 months <br /> on 10/23/84. The Unit was held at 84%
power because of Condenser vacuum problems. On 10/29/84 the Unit was taken below P-8 (36% power) as a precautionary measure to prevent a reactor trip while repressurizing a reactor coolant pump flow transmitter instrument sensing line. Power ascension was resumed on 10/30/84 at 1107 hours0.0128 days <br />0.308 hours <br />0.00183 weeks <br />4.212135e-4 months <br />. Due to continued high back pressure in the Condenser, the Unit was held at 45% power where it remained at the end of the period.
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. REFUELING INFORMATION'
. DOCKET NO.: 50-272' COMPLETED BY: J..Ronafalvy ; UNIT NAME: Salem 1
'DATE: November 10, 1984-
. TELEPHONE: 609/935-6000 EXTENSION: 4455
. Month' October 1984
- 1. . Refueling information has changed-from last' month:
YES NO X
- 2. . Scheduled date for next refueling: February 22, 1986
- 3. Scheduled date for' restart;following refueling: May 4,1986 f
- 4. A) Will. Technical : Specification changes or other license amendments be required?
YES NO NOT DETERMINED TO DATE ~10/1/84 B) - Has the reload fuel design been reviewed by the Station Operating Review Committee?-
YES 'NO X
.If no, w" lien is it scheduled? January 1986-
- 5. Scheduled date (s) for submitting proposed licensing action:
January 1986 if required
- 6. Important licensing considerations associated with refueling:
NONE
- 7. Number of Fuel Assemblies:
A) Incore 193 B) In Spent Fuel Storage 296
- 8. Present licensed spent fuel storage capacity: 1170 I Future spent fuel storage capacity: 1170
{ 9.- Date of last refueling that can be discharged
- to spent fuel pool assuming the present
] licensed capacity: September 2001 r
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4 l 0 Public Service Electric and Gas Company P.O. Box E Hancocks Bridge, New Jersey 08038
. 4 Salem Generating Station '
, November 10, 1984
, A .
Director, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
t MONTHLY OPERATING REFORT S ALEM NO. 1 DOCKET NO. 50-272 ,
In Compliance with Section 6.9, Reporting kequirements for the Salem Technical Specifications, 10 copi~es of the following monthly operating reports for the month of October 1984 are being sent to you.
Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions Major Plant Modification Safety Related Work Orders Operating Summary Refueling Information Sincerely yours, J. M. Zupko, r.
General Manager - Salem Operations JR:sbh cc: Dr. Thomas E. Murley Regional _ Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19406
, Director, Office of Management 2 Information and Program Control U.S. Nuclear Regulatory Commission Washington, DC 20555 l Enclosures Page of l 8-1-7.R4
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i The Eneray People s2m enx n os