ML18096A889
| ML18096A889 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/31/1992 |
| From: | Shedlock M, Vondra C Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9208180090 | |
| Download: ML18096A889 (11) | |
Text
PS~G*
- Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 August 12, 1992 In compliance with Section 6.9.106, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating reports for the month of July 1992 are being sent to you.
RH:pc Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions Safety Related Maintenance 10CFR50.59 Evaluations Operating Summary Refueling Information Sincerely yours, General Manager -
Salem Operations cc:
Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4 The En.erqy Pe_pp~e 9208180090 920731... *-:-~*
PDR ADOCK 0500031r 1
R PDR 95*2189 (10M) 12*89
tllERAGE DAILY UNIT POWER L~
Docket No.:
50-311 Unit Name:
Salem #2 Date:
08/10/92 Completed by:
Mark Shedlock Telephone:
339-2122 Month July 1992 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET)
(MWe-NET) 1 0
17 0
2 0
18 15 3
0 19 233 4
0 20 640 5
0 21 901 6
0 22 883 7
0 23 881 8
0 24 921 9
0 25 1085 10 0
26 1114 11 0
27 1081 12 0
28 1090 13 0
29 1099 14 0
30 1084 15 0
31 1081 16 0
P. 8.1-7 R1
OPERATING DATA REPORT e Docket No:
50-311 Date:
08/10/92 Completed by:
Mark Shedlock Telephone:
339-2122 Operating Status
- 1.
Unit Name Salem No. 2 Notes
- 2.
Reporting Period July:
1992
- 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 NA
- 9.
Power Level to Which Restricted, if any (Net MWe)
N/A
- 10. Reasons for Restrictions, if any ~~~~N~A..:._~~~~~~~~~~~~~~-
- 11. Hours in Reporting Period
- 12. No. of Hrs. Rx. was Critical
- 13. Reactor Reserve Shutdown Hrs.
- 14. Hours Generator On-Line
- 15. Unit Reserve Shutdown Hours
- 16. Gross Thermal Energy Generated (MWH)
- 17. Gross Elec. Energy Generated (MWH)
- 18. Net Elec. Energy Gen. (MWH)
- 19. Unit Service Factor
- 20. Unit Availability Factor
- 21. Unit Capacity Factor (using MDC Net)
- 22. Unit Capacity Factor (using DER Net)
- 23. Unit Forced Outage Rate This Month 744 393.9 0
318.6 0
957403.2 305480 282901 42.8 42.8 34.4 34.1 40.3 Year to Date cumulative 5111 94680 1555.6 60171. 8 0
0 1140.50 58039.3 0
0 3234122.4 133345844.2 1014970 60742018 914312 57782597 22.3 61.3 22.3 61.3 16.2 55.2 16.0 54.7 30.6 23.4
- 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
None
- 25. If shutdown at end of Report Period, Estimated Date of,Startup:
NA 8-1-7.R2
NO.
DATE 0014 07101/92 F
1 2
F:
Forced S:
Scheduled DURATION TYPE 1 (HOURS)
REASON2 425.5 B
Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refueling D-Requlatory Restriction UNIT SHUTDOYN AND POYER REDUCTIONS 4
REPORT MONTH JULY 1992 METHOD OF SHUTTING DOYN REACTOR LICENSE EVENT REPORT #
CH 3
Method:
1-Manual 2-Manual Scram SYSTEM CODE4 E-Operator Training & License Examination F-Administrative 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther G-Operational Error (Explain)
H-Other (Explain)
COMPONENT CODE5 DOCKET NO.
UNIT NAME DATE COMPLETED BY TELEPHONE 50-311 Salem #2 08/10/92 Mark Sbedlock 339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE PIPEXX STEAM GEN. PIPING REPAIRS 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same source I
i I
-c--------- ------
SAFETY RELATED MAINTEN..
E MONTH:
JULY 1992 DOCKET.:
UNIT NAME:
50-311 SALEM 2 WO NO UNIT 920701246 2
920715122 2
920727172 2
9205080.59 2
920605102 2
920605109 2
DATE:
COMPLETED BY:
TELEPHONE:
AUGUST 10, 1992 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION VALVE 24SW243 FAILURE DESCRIPTION:
SW ROOT VALVE INST LINE BREAK -
REPAIR NUCLEAR INSTRUMENT 2NI35D FAILURE DESCRIPTION:
START UP BEZEL READS HIGH -
INVESTIGATE VALVE 21AF8 FAILURE DESCRIPTION:
CORRECT LEAKAGE FROM COVER BOLTS AUX. FEEDWATER STORAGE TANK FAILURE DESCRIPTION:
IMPROPER LEVEL SETPOINT -
INVESTIGATE & CORRECT 21 STEAM GENERATOR LEVEL INDICATION FAILURE DESCRIPTION:
2LI518 4.4% GREATER THAN 2LI640 -
INVESTIGATE 23 STEAM GENERATOR LEVEL INDICATION FAILURE DESCRIPTION:
2LI538 4% GREATER THAN 2LI642 -
INVESTIGATE
10CFR50.59 EVALUATIONS MONTH: -
JULY 1992 e
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 AUGUST 10, 1992 J. FEST (609)339-2904 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFR50.59.
The Station Operations Review Committee has reviewed and concurs with these evaluations.
ITEM
SUMMARY
A.
Design Change Packages (DCP)
DCP# 2EC-3157 Pkg. 1 DCP# 2EC-3158 Pkg. 1 DCP# 2EC-3159 Pkg. 1 "Relocate SJ19 and SJ63 Circuitry (limit switch, solenoid valve, relays) from existing breakers to existing spares at 125 voe distribution cabinets 2AADC and 2CCDC" - This DCP will entail the relocation of non-EQ circuits in the 125 voe Distribution Cabinets 2AADC and 2CCDC to existing spare breakers in the same cabinets.
This proposal does not reduce the margin of safety as defined in the basis for any Technical Specification.
The reliability of the existing circuitry will be improved by load redistribution and Reg. Guide 1.97 compliance will be accomplished.
(SORC 92-073)
"Steam Generator Feedpump High Discharge Pressure Trip" -
The purpose of this DCP is to provide an individual Steam Generator Feed Pump high discharge pressure trip circuit utilizing the existing feed pump discharge pressure transmitters which will allow tripping one feed pump at a time in the event Feed Pump discharge pressure exceeds 1750 psig.
This proposal does not reduce the margin of safety as defined in the basis for any Technical Specification.
The only Technical Specification which involves the Steam Generator Feedpump circuitry involves a trip of the Main feed Feedpump starting the motor driven Auxiliary Feedwater Pump.
This DCP does not alter that function in any way.
(SORC 92-074)
"Redesignation of Breakers 2CV2, 2CV277, 2CV131, 2CV134, 2CV278 Circuitry From Existing Breakers to Available Breakers at 125VDC Distribution Cabinets 2BBDC and 2CCDC" -
The purpose of this design change is to redesignate Breakers 2CV2, 2CV277, 2CV131, 2CV134, 2CV278 circuitry from existing breakers to available breakers at 125VDC Distribution Cabinets 2BBDC and 2CCDC.
iOCFR50.59 EVALUATIONS.
MONTH: -
JULY 1992 (cont'd)
ITEM DCP# 2EC-3160 Pkg. 1 DOCKET.:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 AUGUST 10, 1992 J. FEST (609)339-2904 These changes are being made for the following reason.
During and after a LOCA, loss of Reg.
Guide 1.97 containment isolation valves status indication may result for 2CV3, 2CV5 and 2CV7 due to valve position limit switches sharing a breaker with non-qualified components in containment.
This DCP does not reduce the margin of safety as defined in the basis for any Technical Specification because the reliability of the existing circuitry will be improved by load redistribution and Reg. Guide 1.97 compliance will be accomplished.
(SORC 92-074)
"Isolation of PORV Pressure Switch and Solenoid Circuit" -
The purpose of this change is to separate the PORV (2PR1, 2PR2) auxiliary air supply control circuits from the PORV position indication circuits by making the following modifications:
2AADC -
Removes auxiliary air supply pressure switch 2PD9859 and solenoid valve 25V1198 circuit for PORV 2PR1 from breaker
- 8; reconnects the circuit to circuit breaker
- 40.
Two fuses, one on each leg of circuit breaker #40, are added for proper penetration protection.
Valve 2SJ68 in disconnected from CB
- 41 and reconnected to CB #13 with CB #13 existing load which are all EQ.
This improves reliability of R/G 1.97 indication circuits.
2BBDC -
Removes auxiliary air supply pressure switch 2PD9860 and solenoid valve 25V1199 circuit for PORV 2PR2 from breaker #20; reconnects the circuit to circuit breaker #10.
The existing load on circuit breaker #10 is grouped with load on circuit breaker #7.
A new cable is routed between cabinets 2BBDC and TP25-2 to accomplish this.
Two fuses, one on each leg of circuit breaker #10, are added for proper penetration protection.
This design change does not reduce the margin of safety as defined in the basis for any Technical Specifications because the reliability of the existing circuitry will be improved by load redistribution and Reg. Guide 1.97 compliance will be accomplished.
(SORC 92-079)
10CFR50.59 EVALUATIONS.
MONTH: -
JULY 1992 (cont'd)
ITEM DCP# 2EC-3109 Pkg. 1 B.
Temporary Modifications TMR# 92-042 DOCKET":
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 AUGUST 10, 1992 J. FEST (609)339-2904 "Aux. Feedwater Pump No. 23 Digital Tachometer and Temperature Indication Upgrade" -
The purpose of this change is to ensure Salem Station design requirements are fulfilled for the installation of permanent instrumentation for Unit 2 AFP No. 23 speed and discharge piping temperature.
In addition, permanent remote indication of AFP No. 23 turbine inlet steam trap drain line temperature will be provided to eliminate the need for operations personnel entry into a high noise/humidity environment to obtain Surveillance Procedure mandated (as well as INPO recommended) AFP performance parameter measurements.
This proposal will not affect the operability, availability, or capacity of AFP No. 23 or affect the flow path to any Steam Generator as described in the Technical Specifications.
Therefore this change does not reduce the margin of safety as defined in the basis for any Technical Specification.
( SORC 92-083)
"Enhance Leaktightness of Air-Operator Diaphragms for Valves 2PR1, 2PR2, 2PS1, 2PS3" -
The air-operator for valves 2PR1, 2PR2, 2PS1, and 2PS3 will be temporarily modified by installing an 0-ring.
The o-ring will act as a redundant seal to the existing diaphragm.
The sizing and material selection of the 0-ring will be compatible with the actuator geometry and existing diaphragm material.
The o-ring is fabricated from Buna-N rubber.
The material properties of Buna-N rubber are comparable to the EPDM diaphragm.
No adverse interaction between the o-ring and diaphragm is anticipated based on similar material properties.
The dimensions of the o-ring are sized such that no interferences with the actuator movement will occur.
Lab testing has been performed at elevated temperatures that have been postulated to occur in the installed location.
No failure modes resulted from the installation of the o-ring at these temperatures.
Degradation of
10CFR50.59 EVALUATIONS MONTH: -
JULY 1992 (cont'd)
ITEM
- c.
Procedures and Revisions Sl.OP-SO.RC-OOOS(Q)
DOCKET'°:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 AUGUST 10, 1992 J. FEST (609)339-2904 the materials selected can occur if temperatures exceed those anticipated in the installed condition over time.
Complete degradation of the o-ring material is not anticipated for the duration of the TMOD No adverse interaction between the o-ring and diaphragm is expected based on similar material chemistries for rubber elastomers.
Additionally, a lock washer is being installed on each bolt to minimize the potential for loosening of the diaphragm enclosure nuts.
The proposed modification does not reduce the margin of safety as defined in the basis of any Technical Specification.
(SORC 92-079)
"Draining the RCS" Rev. 2 -
The purpose of this procedure is to provide instructions to: 1) drain the Reactor Coolant system (RCS) in preparation for Reactor Vessel Head Removal or Reinstallation; 2) drain the RCS to any desired level including Reduced Inventory Operation or Mid-Loop Operation.
There is no Technical Specification defining either the RCS level or the margin needed to prevent vortex formation when using the RHR pumps.
Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
(SORC 92-084)
SALEM UNIT NO. 2 SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 2 JULY 1992 The Unit began the period shutdown to test feedwater piping wall thickness due to erosion/corrosion concerns identified on Unit 1.
While the Unit was shutdown, a number of design changes were implemented to address Reg. Guide 1.97 issues.
Also during this period, the failed 2"C" 460V transformer was replaced.
The Unit was synchronized on July 18, 1992.
The subsequent power increase was stopped at 90% to perform repairs to No. 22 Heater Drain Pump and clean condenser water boxes.
On July 24, 1992, the Unit was returned to full power and continued to operate at full power throughout the remainder of the period.
.REFUELING INFORMATION MONTH: -
JULY 1992 MONTH JULY 1992 DOCKET'°:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
- 1.
Refueling information has changed from last month:
YES 00 X
- 2.
Scheduled date for next refueling:
MARCH 27, 1993 50-311 SALEM 2 AUGUST 10, 1992 J. FEST (609)339-2904
- 3.
Scheduled date for restart following refueling:
MAY 21, 1993
- 4.
a)
Will Technical Specification changes or other license amendments be required?:
YES NO
_NOT DETERMINED TO DATE ~=x~-
b)
Has the reload fuel design been reviewed by the station Operating Review Committee?:
YES NO x
If no, when is it scheduled?: FEBRUARY 1993
- 5.
Scheduled date(s) for submitting proposed licensing action:
N/A
- 6.
Important licensing considerations associated with refueling:
- 7.
Number of Fuel Assemblies:
- a.
Incore
- b.
In Spent Fuel Storage
- 8.
Present licensed spent fuel storage capacity:
Future spent fuel storage capacity:
193 408 1170 1170
- 9.
Date of last refueling that can be discharged to the spent fuel pool -assuming the present licensed capacity:
March 2003 8-1-7.R4