ML20211E107

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Monthly Operating Rept for Jan 1987
ML20211E107
Person / Time
Site: Salem PSEG icon.png
Issue date: 01/31/1987
From: White P, Zupko J
Public Service Enterprise Group
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
NUDOCS 8702240234
Download: ML20211E107 (15)
v • d • e


Text

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AVERAGE DAILY UNIT POWER LEVEL Docket No. 50-311 Unit Name Salem # 2 Date February 10,1987 Completed by Pell White- Telephone 609-935-6000.

Extension 4451 Month January 1987 Day' Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 986 17 969 2 1068 18 685 3 618 19 285 4 854 20 1043 5 1048 21 1068 6 1054 22 1080 7 1008 23 1098 8 1016 24 1081 9 1014 25 1110 10 1004 26 1108 11 1018 27 1108 12 1021 28 1103 13 957 29 1112 14 1010 30 1104 15 1057 31 1085 16 968 Pg. 8.1-7 R1 i

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OPERATING DATA REPORT Docket-No.-50-311 Date ' February 10,1987 i Telephone 935-6000 Completed by Pell White Extension 4451 Operating Status

1. Unit Name Salem No. 2 Notes
2. Reporting Period January 1987
3. Licensed Thermal Power (MWt) 3411 4.. Nameplate Rating (Gross MWe) 1170
5. Design Electrical Rating (Net MWe) 1115
6. Maximum Dependable Capacity (Gross MWe)1149

-7. Maximum Dependable Capacity (Net MWe) 1106

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) NA
10. Reasons for Restrictions, if any

. This Month Year to Date Cumulative 1

11. Hours in Reporting Period 744 774 46489
12. No. of Hrs. Reactor was Critical 728.8 728.8 26668.7
13. Reactor Reserve Shutdown Hrs. 0 0 0 3 14. Hours Generator On-Line 725.7 725.7 25664.1

.15. Unit Reserve Shutdown Hours 0 0- 0 '

16. Gross Thermal Energy Generated 2346533 2346533

.(MWH) 28355005 ,

17. Gross Elec. Energy Gencrated (MWH) 767530 767530 25944670
18. Net Elec. Energy Generated (MWH) 737725 737725 24585130
19. Unit Service Factor 97.5 97.5 55.2 7
20. Unit Availability Factor 97.5 97.5 55.2
21. Unit Capacity Factor (using MDC Net) 89.7 89.7 47.8
22. Unit Capacity Factor (using DER Net) 88.9 88.9 47.4
23. Unit Forced Outage Rate 2.5 2.5 34.4
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:

N/A

j. 26. Units in Test Status (Prior to Commarcial Operation):

1 Forecast Achieved Initial Criticality 6/30/80 8/2/80 l Initial Electricity 9/1/80 6/3/81 Commercial Operation 9/24/80 10/13/81 8-1-7.R2

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UNIT SHUTDOWN AND POWER REDUCTIONS Docket No. 50-311 .

REPORT MONTH JANUARY 1987 Unit Nsne Salen No.2 Date February 10,1987 Completed by Pell White Telephone 609-935-6000 -

Extension 4451 l

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Method of i Duration Shutting License Cause and Corrective

No. Date Type Hours Reason Down Event System Component Action to 1 2 Reactor Report Code 4 Code 5 Prevent Recurrence J

Station Service Transformer Station j 0002 12-31-86 F 9.75 B 5 ----

EC TRANSF Power Auxiliary Other Feedwater 0016 1-2-87 F .47 B 5 ----

HH PUMPXX Pump Problems Other Feedwater 0018 1-3-87 F 13.03 B 5 ----

HH PUMPXX Pump Problems

Other Feedwater j 0020 1-3-87 F 8.58 B 5 ----

HH PUMPXX Pump Problems Other Circulating j Water System 0022 1-3-87 F 6.67 A 5 ----

HF PUMPXX Problems Station Service 1 Transformer Station 0024 1-4-87 F 8.83 B 5 ----

EG TRANSF Power Auxiliary Station Service Transformer Station 0026 1-4-87 F 10.50 B 5 ----

EG TRANSF Power Auxiliary Nuclear Residual

Heat Removal / Delay i 0042 1-3-87 F 1.48 D 5 ---- CF PIPEXX Heat Removal Reactor Control 0058 1-18-87 F 18.32 G 3 ----

IB INSTRU Systems i

1 2 Reason 3 Method 4 Exhibit G 5 Exhibit 1 F: Forced A-Equipment Failure-explain 1-Manual Instructions Salem as S: Scheduled B-Maintenance or Test 2-Manual Scram. for Prepara- Source C-Refueling 3-Automatic Scram, tion of Data D-Regulatory Restrictiou 4-Continuation of Entry Sheets E-Operator Training & Licensing Exam Previous Outage for Licensee F-Administrative 5-Load Reduction Event Report G-Operational Error-explain 9-Other (LER) File H-Other-explain (NUREG 0161)

MAJOR PLANT MODIFICATIONS DOCKET NO.: 50-311 REPORT MONTH JANUARY 1987 UNIT NAME: Salem 2 DATE: February 10, 1987 COMPLETED BY: L. Miller TELEPHONE: _609/339-4497

  • DCR NO. PRINCIPAL SYSTEM SUBJECT 2EC-0641 Fuel Handling Install cable support system to supplement existing stop log removal cable system.

2EC-1439 Pressurizer System The control for the emergency feed breaker does not provide a flashing green light and 4

associated audible indication

, when the breaker trips due to over current or a local trip.

(breaker for the "2GP" pressurizer heater bus and "2EP" pressurizer heater bus) add flashing and audible features to breaker trip indication.

2EC-1563 Fire Protection Wrap "B" diesel control cable tray 1A217 between tray 1A258 and 1A218 with ES195, one hour Fire Barrier material.

2EC-2061 Containment Airlock Replace the teflon packing in the shaft seal assemblies of the Containment Airlocks with a Grafoil packing.

2SC-0348 Component Cooling Provide Motor operated valvec System for the auxiliary header

isolation valves 2CC30 and 2CC31.

I 2SC-1531 Containment Ventil. Shorten the section of duct i

' work that protrudes from the containment wall that is directly over the containment equipment hatch stand.

1 Design Change Request

MAJOR PLANT MODIFICATIONS DOCKET NO: 50-311 REPORT MONTH - JANUARY 1986 UNIT NAME: SALEM 2 DATE: February 10,1987 COMPLETED BY: L. Miller TELEPHONE: (609)339-4497

  • DCR SAFETY EVALUATION 10 CFR 50.59 2EC-0641 This DCR provides additional supports to the monorail system on Elevation 130' in the Fuel Handling Building. The additional supports increase the margin of safety by increasing the capacity of the monorail to withstand impact loads associated with the safety cable support system should a hoist failure occur. The modification is on non-safety related equipment located in a Q Area. The seismic installation of the monorail is not affected as the supports meet the same seismic design criteria. The seismic mounting design precludes the possibility of physical damage to any safety related equipment in the area. No new failure modes were introduced by this modification. This design change upgraded the capability of existing equipment. There is no impact on any analysis in the UFSAR or the Technical Specifications. This modification does not alter any plant process or discharge or affect the environmental impact of the plant. No unreviewed safety or environmental questions are involved.

2EC-1439 This change added an additional signal for breaker trip. This modification does not affect any safety related system and does not degrade any safety related equipment or circuit. This change did not alter any plant process or discharge or affect the environmental impact of the plant. No unreviewed safety or environmental questions are involved.

2EC-1563 This design change provided additional protection by wrapping the cable tray with one hour fire barrier material. This modification did not alter any plant process or discharge or affect the environmental impact of the plant. No reviewed safety or environmental questions are involved.

  • DCR - Design Change Request

MAJOR PLANT MODIFICATIONS DOCKET NO: 50-311 REPORT MONTH - JANUARY 1986 UNIT NAME: SALEM 2 DATE: February 10,1987 COMPLETED BY: L. Miller TELEPHONE: (609)339-4497

  • DCR SAFETY EVALUATION 10 CFR 50.59 2EC-2061 The grafoil packing installed by this design change is an acceptable alternative to the teflon packing for its function in the containment air locks. The total radiation dose seen in the contianment for forty years life and a design bassis accident is significantly less than that required to result in a 25% damage dose to the grafoil packing. Because Grafoil has a much greater resistance to radiation than teflon, the use of grafoil packing enhances the margin of safety on the shaft and seal. This modification does not alter any plant process or discharge and does not affect the environmental impact of the plant. No unreviewed safety or environmental questions are involved.

2SC-0348 This change did not involve any change to the UFSAR or Technical Specifications. The valve function remains the same. The change from manual to motor operated valves does not alter the function of the system. There are no changes to station effluent releases or the environmental impact of the plant. No unreviewed safety or environmental questions are involved.

2SC-1531 The goose-neck duct carries 4% of the total airflow of the containment ventilation system.

Removal of the goose-neck changes the airflow from slightly downward to horizontal. This portion of airflow eventually mixes with the remaining airflow and provides cooling without any reduction in the cooling capacity of the system. Removal of the goose-neck reduced the weight of the duct system. This modification has no impact on any analysis performed in the UFSAR or to any basis to the Technical Specifications.

Thee is no chance of having created a new malfunction or aggravating a previous one. This change did not alter any plant process or discharge or affect the environmental impact of the plant. No unreviewed safety or environmental questions are involved.

Design Change Request

PSE&G SALEM GENERATING STATION

  • SAFETY RELATED WORK ORDER LOG .

SALEM UNIT 2 WO NO UNIT EQUIPMENT IDENTIFICATION 0099171180 2 REACTOR TRIP BREAKERS FAILURE DESCRIPTION: PERFORM REACTOR TRIP BREAKER SURVEILLANCE PROCEDURES 24 HOURS PRIOR TO START-UP. 2IC-18.1.006, 2IC-18.1.007, 2IC-18.1.010, 2IC-18.1.011.

CORRECTIVE ACTION: PERFORMED 2IC-18.1.006, 18.1.007, 18.1.010 AND 18.1.011 TRIP BREAK A UU=7 CYCLES TRIP BREAKER A SHUNT = 6 CYCLES, TRIP BREAKER B UU = 7 CYLES, TRIP BREAKER B SHUNT = 6 CYCLES.

0099186985 2 START AIR RECEIVER CHECK VALVE FAILURE DESCRIPTION: 2B DIESEL GENERATOR STARTING AIR RECEIVER CHECK VALVE LEAK THRU. PLEASE REPAIR.

CORRECTIVE ACTION: REPLACED SPRING AND GASKET, LAPPED PLUG TO SEAT, BLUE CHECK, SAT.

0099187299 2 ESF FAILURE DESCRIPTION: TRAIN "A" STEAMLINE PRESSURE SI DOES NOT INDICATE

" UNBLOCKED" ON CONSOLE.'RP4 INDICATIONS INDICATE NORMAL " UNBLOCKED" CONDITION.

CORRECTIVE ACTION: REPLACED A710 MEMORY COIL SN 0001 WITH SN 1378-PERFORMED NECESSARY STEP OF IC-18.1.008 AND VERIFIED NO INTERMITTENT CONDITION. ALSO VERIFIED STEAMLINE PRESSURE BLOCK CLEARED.

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SALEM UNIT 2 -

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WO NO UNIT EQUIPMENT IDENTIFICATION 8511280014 2 FCU OUTLET VALVE FAILURE DESCRIPTION: THE VALVE WAS CLOSED FROM THE CONTROL ROOM AND WILL'

NOT REOPEN. PLEASE INVESTIGATE AND' REPAIR.

I CORRECTIVE ACTION: ADJUSTED VALVE. LINKAGE AND LIMIT SWITCH. VERIFIED

CORRECT OPERATION WITH CONTROL.

i 8608230054

) 2 PERF, VT MECH. SNUBBERS FAILURE DESCRIPTION: PERFORM A VISUAL EXAMINATION ON 100% OF THE MECHANICAL.

i SNUBBERS IN ACCORDANCE WITH PROCEDURE M9-IIP-1.

j CORRECTIVE ACTION: ALL MECHANICAL SNUBBERS (11) WERE VISUALLY EXAMINED IN ACCORDANCE WITH PROCEDURE M9-IIP-1 WITH SATISFACTORY RESULTS.

1 I 8608140829 i 2 SERVICE WATER ,

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l FAILURE DESCRIPTION: REPLACE EXISTING C.S. PIPING WITH S.S. IN THE i

) DISCHARGE OFF THE 21 NUCLEAR HEADER UPSTREAM AND j

DOWNSTREAM OF 21SW49. INSTALL FLANGES UPSTREAM OF r THE TIE-IN TO THE DISCHARGE LINE. '

i j CORRECTIVE ACTION: REPLACED C.S. PIPING WITH S.S. UPSTREAM, AND  ;

j DOWNSTREAM OF 21SW49. INSTALLED 24" FLANGES UPSTREAM i i

OF THE TIE-IN TO DISCHARGE LINE.

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SALEM UNIT 2 -

WO NO UNIT EQUIPMENT IDENTIFICATION 1

8608070065 2 24SW STRAINER FAILURE DESCRIPTION: 24SW PUMP STRAINER IS BREAKING SHEAR KEYS. PLEASE REPAIR.

CORRECTIVE ACTION: DISASSEMBLED STRAINER, REPLACED ELEMENTS, REASSEMBLED, REPLACED SHEAR KEY, REPACKED SAME, l REWIRED MOTOR.

8611090161 2 22SW91

} FAILURE DESCRIPTION: NO STEM POSITION INDICATOR.

CORRECTIVE ACTION: VALVE STEM INSTALLED, CAUTION TAG ON VALVE TO ADJUST

{ STEM NUTS WHENEVER VALVE IS SHUT, VALVE IS OPERABLE.  !

! 8611110030 j 2 MSIV HYDRAULIC SNUBBER FAILURE DESCRIPTION: REPLACE THE O-RING BETWEEN THE 90 DEGREE ELBOW FROM

  1. THE MAKE-UP RESERVOIR SUPPLY PIPING TO THE CONTROL VALVE ASSEMBLY. VISUALLY INSPECT FOR LEAKS AFTER
RESERVOIR HAS BEEN FILLED.

I CORRECTIVE ACTION: INSTALLED NEW 0-RING BETWEEN THE 90 DEGREE ELBOW-FROM THE MAKE-UP RESERVOIR SUPPLY PIPING TO THE CONTROL VALVE ASSEMBLY. VISUAL INSPECTION REVEALED NO LEAKS.

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SALEM UNIT'2

  • WO NO UNIT EQUIPMENT IDENTIFICATION j 8611170547 2 CS PP RM COOLER CONTROL FAILURE DESCRIPTION: VALVE IS STUCK HALFWAY OPEN, PLEASE. REPAIR. VALVE i NEEDED, PERFORM 4.0.5-V.

l CORRECTIVE ACTION: REMOVED VALVE BONNET, STEM WAS NOT BENT, REMOVED

. DEBRIS FROM PIPING, REPACKED VALVE WITH CHESTERTON

PACKING AND REINSTALLED VALVE, VALVE IN OPEN i POSITION.

i 8611300157 2 22 CCHX CONTROL 2

FAILURE DESCRIPTION: THE 22SW127 WILL NOT CLOSE ALL THE WAY. IS STILL i

OPEN 1/8 OF THE WAY. PLEASE VERIFY THE STROKE IS CORRECT.

CORRECTIVE ACTION: FOUND VALVE NOT TRAVELING TO FULL CLOSED POSITION.

REMOVED VANE ARM OFF SHAFT AND SET BACK OF TOOTH TO GET FULL STROKE. STROKED VALVE THRU POSITIONER TO VERIFY PROPER OPERATION. r 1

i 8612040876 2 2B 125 VDC GROUND FAILURE DESCRIPTION: 2B 125 VDC BATTERY BUS HAS A GROUND INDICATED. THE GROUND HAS BEEN ISOLATED TO BREAKER 2BDC2AX29, 22 CFCU FILTER DAMPERS. PLEASE INVESTIGATE AND REPAIR.

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J CORRECTIVE ACTION: FOUND WATER IN CONDUIT ON 22 CFCU DRAIN WATER,

DRIED OUT.

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SALEM UNIT 2 -

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, WO NO UNIT EQUIPMENT IDENTIFICATION ,

j 8612070082 2 22 W.G. COMP.

I FAILURE DESCRIPTION: 22 W.G. COMP. NO DISCHARGE PRESSURE, PLEASE INVESTIGATE AND REPAIR.

CORRECTIVE ACTION: REPLACED DIAPHRAGM ON 22WG10.

8612120063 2 22 RCP SEAL WATER TEMP i FAILURE DESCRIPTION: INLET TEMPERATURE INDICATION FOR SEAL WATER TO 22 RCP IS INDICATING ABNORMALLY LOW CN THE CONSOLE.

I COMPUTER INDICATION IS NORMAL. PLEASE INVESTIGATE j AND REPAIR.

CORRECTIVE ACTION: VERIFIED THERMOCOUPLE READING LOW IN RACK WITH TEMP-FLUKE. FOUND THERMOCOUPLE FOR 63 DEG F. BEZEL INDICATOR SWAPPED. REVERSED LEADS AND VERIFIED i PROPER INDICATION. BEZEL 79 DEG F.. COMPUTER

81 DEG. F.

i 4 8612130051

] 2 CONTAINMENT AIRLOCK'

} FAILURE DESCRIPTION: 100 ELEV. CONT. AIRLOCK OUTSIDE DOOR AS A BOLT

MISSING IN THE CLOSING MECHANISM. THIS RESULTS IN l SLOPPY OPERATION OF AIRLOCK DOOR. PLEASE REPAIR

.j BEFORE DOOR BECOMES DAMAGED.

CORRECTIVE ACTION: REPLACED BOLT AND ADJUSTED DOOR SWING STOP MECHANISM.

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s SALEM UNIT 2

  • WO NO UNIT EQUIPMENT IDENTIFICATION 8612160031 2 REACTOR HEAD VENT FAILURE DESCRIPTION: 2RC900 HAS A SMALL PACKING LEAK. WAS UNABLE TO TAKE UP ON PACKING, BECAUSE OF VALVE LOCATION. PLEASE REPAIR.

CORRECTIVE ACTION: PULLED UP ON PACKING, STOPPED LEAK.

8612200016 2 24 CFCU FAILURE DESCRIPTION: NO SW WATER FLOW TO FCU WHEN PLACED IN SERVICE.

PLEASE INVESTIGATE AND REPAIR.

CORRECTIVE ACTION: FOUND FEEDBACK LINKAGE DISCONNECTED ON 24SW57.

RECONNECTED LINKAGE AND VERIFIED FLOW IN HIGH AND LOW SPEED WITH CONTROL ROOM OPERATOR.

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT NO. 2 JANUARY 1987 The Unit began-the period operating at one hundred perecent (100%)

power. The electrical output was limited to eleven hundred (1100) MW gross due to equipment being removed from service because of station i power transformer limitations.

On January 2, 1987, power was reduced because No. 21 Steam Generator Feed Pump tripped on low suction pressure. Excess grass at the circulating water screens impaired the circulating water flow and affected the condensing capabilities which caused a reduction in condensate flow. The grass conditions improved and No. 21 Steam Generator Feed Pump was returned to service.

! The Unit was returned to full power on January 4, 1967. The power l'

level was held at ninety five percent (95%) until the control circuitry on the feed pumps could be recalibrated.

On January 13, 1937, it was discovered that two of four RHR cold leg injection lines were isolated to perform preventive maintenance, the plant configurationn was similar to that described in a NRC Inforamtion Notice 87-01 In response to an interpretation of the notice, a shutdown was initiated to assure compliance with Technical Specifications. The shutdown was terminated at sixty six percent i (66%) power, when the RHR loops were returned to service, restoring normal plant configuration.

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_ Power was increased to one hundred percent (100%) on January 15, 1987, following completion of feedwater instrumentation calibration. On

January 16, 1987, a low system demand resulted in a load reduction to i eighty percent (80%).
On January 18, 1987, the turbine generator was taken off line due to l indication of an exciter ground fault. A nuclear instrument
intermediate range channel rod block indication appeared while the reaactor was still critical and in the process of correcting the
problem, the reactor tripped. The problem was corrected and the Unit
returned to service on January 19, 1987.

! The Unit operated at one hundred percent (100%) power for the i' remainder of the period.

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REFUELING INFORMATION

. DOCKET NO.: 50-311 COMPLETED BY: L.K. Miller UNIT NAME: Salem 2 DATE: February 10, 1987-TELEPHONE: 609/935-6000 EXTENSION: 4497 Month January 1987

1. Refueling information has changed from last month:

YES X NO

2. Scheduled date for next refueling: April 9, 1988 ,
3. Scheduled date for restart following refueling:

June 8, 1988

4. A) Will Technical Specification changes or other license amendments be required?

YES NO X B) Hr.s the reload fuel design been reviewed by the Station Operating Review Committee?

YES NO X If no, when is it scheduled? February, 1988

5. Scheduled date(s) for submitting proposed licensing action:

February 1988 if required

6. Important licensing considerations associated with refueling:

NONE

7. Number of Fuel Assemblies:

A) Incore 193 B) In Spent Fuel Storage 224

8. Present licensed spent fuel storage capacity: 1170 Future spent fuel storage capacity: 1170
9. Date of last refueling that can be discharged to spent fuel pool assuming the present licensed capacity: March 2003 8-1-7.R4

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O PSIEG Public Service Electric and Gas Company P.O. Box E Hancocks Bridge, New Jersey 08038 Salern Generating Station February 17, 1987 Director, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Sir:

MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 In compliance with Section 6.9, Reporting Requirements for the Salem Technical Specification, 10 copies of the following monthly operating reports for the month of January 1987 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, Uf . fo/l /

J. M. Zupko, Jr.

General Manager - Salem Operations JR:sl cc: Dr. Thomas E. Murley Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19406 Director, Office of Management U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Enclosures 8-1-7.R4 6}J

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