ML18096A888

From kanterella
Jump to navigation Jump to search
Monthly Operating Rept for Jul 1992 for Salem Unit 1.W/ 920812 Ltr
ML18096A888
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/31/1992
From: Shedlock M, Vondra C
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9208180084
Download: ML18096A888 (15)
v • d • e


Text

PS~*

Public Service Electric arid Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 12, 1992 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

MONTHLY OPERATING REPORT SALEM NO. 1 DOCKET NO. 50-272 In compliance with Section 6.9.1.6, 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.

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,

-~~UV~

General Manager -

Salem Operations RH:pc cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-l-7.R4 r Ihe_En_erav People 9200100004** 920731 -- ~ :~*~

10 2 4 PDR ADOCK 05000272 95-2189 (10M) 12-89 R PDR

AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-272 Unit Name: Salem #1 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 0 3 0 19 0 4 0 20 0 5 0 21 0 6 0 22 0 7 0 23 0 8 0 24 0 9 0 25 0 io 0 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14 0 30 0 15 0 31 0 16 0 P. 8.1-7 R1

  • t OPERATING DATA REPORT Docket No: 50-272 Date: 08/10/92 Completed by: Mark Shedlock Telephone: 339-2122 Operating Status
1. Unit Name Salem No. 1 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=--~~~~~~~~~~~

This Month Year to Date Cumulative

11. Hours in Reporting Period 744 5111 132264
12. No. of Hrs. Rx. was Critical 0 2256.2 85856.4
13. Reactor Reserve Shutdown Hrs. 0 0 0
14. Hours Generator On-Line 0 2117.8 83165.7
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 0 7203120.0 262700265.2
17. Gross Elec. Energy Generated (MWH) 0 2416240 87220870
18. Net Elec. Energy Gen. (MWH) -6738 2299134 83075707
19. Unit Service Factor 0 41.4 62.9
20. Unit Availability Factor 0 41.4 62.9
21. Unit Capacity Factor (using MDC Net) 0 40.7 56.8
22. Unit Capacity Factor (using DER Net) 0 40.3 56.3
23. Unit Forced Outage Rate 100 30.5 21.6
24. Shutdowns scheduled over next 6 months (type, date and duration of each)

We are presently in a forced maintenance outage.

25. If shutdown at end of Report Period, Estimated Date of startup:

August 13. 1992 8-l-7.R2

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH JULY 1992 DOCKET NO. 50-272 UNIT NAME Salem #1 DATE 08/10/92 COMPLETED BY Mark Shedlock '

TELEPHONE 339-2122 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION 1 2 NO. DATE TYPE (HOURS) REASON REACTOR REPORT # CODE 4 CODE 5 TO PREVENT RECURRENCE 0017 07/01/92 F 744 B 4 ------ CH PIPEXX STEAM GEN. PIPING REPAIRS 1 2 3 4 5 F: Forced Reason Method: Exhibit G - Instructions Exhibit 1 - Same S: Scheduled A-Equipment Failure (explain) 1-Manual for Preparation of Data Source B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling 3-Automatic Scram Event Report CLER) File D-Requlatory Restriction 4-Continuation of CNUREG-0161)

E-Operator Training & License Examination Previous Outage F-Administrative 5-Load Reduction G-Operational Error (Explain) 9-0ther H-Other (Explain)

SAFETY RELATED MAINTENANCE DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 WO NO UNIT EQUIPMENT IDENTIFICATION 920507108 1 #2 STATION AIR COMPRESSOR FAILURE DESCRIPTION: AFTERCOOLER LEAKING SERVICE WATER

- REPAIR 910409128 1 VALVE 1WL99 FAILURE DESCRIPTION: 1WL99 RCDT NOT SEATING -

TROUBLESHOOT 910429147 1 VALVE 1PR6 FAILURE DESCRIPTION: 1PR6 WIRING DEGRADED - REPLACE 920128152 1 VALVE 1SWE28 FAILURE DESCRIPTION: LEAK AT COVER - REMOVE AND INSPECT 920212194 1 CHARGING PUMPS FAILURE DESCRIPTION: EXHAUST VENTILATION LOW AIR FLOW

- INVESTIGATE 920217088 1 15 SERVICE WATER PUMP FAILURE DESCRIPTION: PUMP FAILED 4.0.5P FULL FLOW TESTING - INVESTIGATE 920601186 1 11 RHR PUMP FAILURE DESCRIPTION: SEAL LEAKS - REWORK AS REQUIRED 920608126 1 VALVE 12SW379 FAILURE DESCRIPTION: AIR LEAK ON INSTRUMENT LINE TO 12SW379 - REPAIR 920608189 1 12 COMPONENT COOLING HEAT EXCHANGER FAILURE DESCRIPTION: TEMPERATURE CONTROL FAULTY -

TROUBLESHOOT 920611236 1 VALVE 14SW65 FAILURE DESCRIPTION: VALVE HAS GASKET LEAK - REPLACE GASKET

SAFETY RELATED MAINTENANCE DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST (cont'd) TELEPHONE: (609)339-2904 WO NO UNIT EQUIPMENT IDENTIFICATION 920618107 1 VALVE 1CC215 FAILURE DESCRIPTION: VALVE HAS NO OPEN LIMIT INDICATION - INVESTIGATE 920710099 1 VALVE 1SJ4 FAILURE DESCRIPTION: TORQUE SET HIGH - TROUBLESHOOT &

REWORK AS REQUIRED 920728112 1 VALVE 11VC17 FAILURE DESCRIPTION: NO LIMIT SWITCH INDICATION -

TROUBLESHOOT 920728155 1 RADIATION MONITOR 1R12A FAILURE DESCRIPTION: REPLACE 1R12A

10CFR50.59,EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (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# lEX-2285 Pkg. 1 "Full Flow Test - Aux. Feedwater" - This special test is the performance of a full flow Auxiliary Feedwater (AFW) System test to measure maximum AFW flow rates and their associated pressures. The test data will be used to benchmark and validate the current analytical model (RETRAN) for the AFW System. The validation of the RETRAN model is necessary to establish a baseline for the existing system, and ensure proper modification of the trim for flow control valves 11-14AF11 and ll-14AF21.

The overall objective of the project is to reduce the capacity of the AFW System by modifying the flow characteristics of AFll and AF21 valves. The reduction in the AFW flow delivery capability is essential to avoid exceeding the containment safety limits, under certain accident conditions, for the future fuel cycles. Other concerns that will be addressed by implementing AFW flow reduction include RCS over-cooling following a reactor trip, increased potential for turbine driven pump cavitation, and the current need to throttle AF21s beyond their optimum design throttle range. (SORC 92-072)

DCP# lE0-2284 Pkg. 1 "Replacement of Feedwater Pipe-To-Steam Generator Nozzle Fittings" - The purpose of this change is to document the replacement of the Nos. 11, 12, 13, and 14 feedwater pipe fittings adjacent to the steam generators. The subject fittings are being replaced to disposition the nonconforming conditions documented in DR Nos. SMD 92-681, SMD 92-682, SMD 92-690, SMD 92-692 and SMD 92-693.

The new fittings are functionally identical to the existing fittings except that the interface fittings at the No. 13 and 14 steam generators do not accommodate the installation of the exploratory thermocouples mounted in the current fittings. The subject thermocouples are abandoned in place and are not shown on any functional drawings. They are however depicted on the

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

~------------------------------------------------------------

construction piping isometrics which will be updated via this Configuration Change Package.

Additionally, it was determined that the concentric reducers currently installed on the Nos. 11 and 12 steam generators were fabricated from ASME SA 105 material versus ASTM A 234 Grade WPC as required by the piping specification.

Although SA 105 is comparable to A 234 Grade WPC, the use of SA 105 fitting was not reflected in the plant documentation. The replacement fittings, which are fabricated from ASME SA 234 Grade WPC, will restore the plant to its documented configuration. The counterbore details for all of the new fittings have been modified to accommodate ISI. All affected documentation will be updated via this package. The replacement of the fittings and welds will restore the plant to its analyzed configuration. Therefore, the replacement of the fittings does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 92-072)

DCP# lEC-3185 Pkg. 1 "Reg. Guide 1.97 Modifications for 1SJ19, 1SJ63, and 1SJ68" - The purpose of this change is to relocate SJ19 circuitry (including the 74/DC relay) from breaker 1CCDC20 to existing spare 1CCDC22 at the 125VDC Distribution Cabinet.

Relocate the SJ63 circuitry from breaker 1AADC13 to 1AADC41 to relocate SJ68 circuitry from 1AACD41 to 1AADC13. 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)

DCP# lEC-3187 Pkg. 1 "Redesignating of Non-EQ Circuits in 125 VDC Distribution Cabinets" - The purpose of this change entails redesignating of Non-EQ circuits in the 125 VDC Distribution Cabinets lBBDC and lCCDC to their own breakers in the same cabinets. These cabinets are located in the Unit 1 Auxiliary Building, Control Area Elevation 100' - 0 11

  • Circuitry for 1CV2 and 1CV277 is not

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

environmentally qualified and is therefore being separated from 1CV3 and 1CV7 which are qualified and required for Reg. Guide 1.97 indication.

Similar circuitry for 1CV131, 1CV134 and 1CV278 is not qualified for Reg. Guide 1.97 indication.

Limit switches associated with the EQ (1CV3) and Non-EQ (1CV2, 1CV277) solenoids will also be separated inside containment for the "B" Bus in Junction Box EJT-210. Separation will be accomplished utilizing existing spare conductors.

These spares however will require Butt Splicing-in the currently fire wrapped electrical penetration 1-41. In 1BBDC, the 1CV2 and 1CV277 circuitry will be removed from breaker #19 (FU19/FU19A) and will be connected to it's own Breaker #22. There is no impact to the existing load (1SJ19) currently-on Breaker #22. In order to add new fuses, the front of the panel will require a small cutout with a stainless steel door, connected by a piano hinge. The two required fuses will be designated as FU22B and FU22C. This change will provide two enhancements to the circuitry. First, Non-EQ circuitry will be separated from EQ circuitry to preclude the possible loss of Reg Guide 1.97 indication during and after a LOCA.

Second, the existing circuit breaker and branch cable loading will be reduced by distributing load to the new breaker. In separating the EQ and Non-EQ components, the loss of voltage relay remains with the power supply for the EQ devices.

The power source for the Non-EQ components will have a new loss of 125 VDC alarm relay

("Struther's Dunn Model #219ABA152") equivalent to the existing relays installed with the EQ power supply. Annunciator points will be reflected in the "Auxiliary Alarm Typewriter List". 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 92-080)

DCP# 1EC-3188 Pkg. 1 "Isolation of PORV Pressure Switch and Solenoid Circuit" - The purpose of this change is to separate the PORV (1PR1, 1PR2) auxiliary air

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

supply control circuits from the PORV position indication circuits by making the following modifications: lAADC - Removes auxiliary air supply pressure switch 1PD9859 and solenoid valve 1SV1198 circuit for PORV lPRl from breaker #8; reconnects the circuit to circuit breaker #40.

Two fuses, one on each leg of circuit breaker #40, are being used for proper penetration protection.

A loss of power relay (74) is added for monitoring DC power supply. A new cable is routed between cabinets lAADC and TP14-2. lBBDC - Removes auxiliary air supply pressure switch 1PD9860 and solenoid valve 1SV1199 circuit for PORV 1PR2 from breaker #20; reconnects the circuit to circuit breaker #10. The existing load on circuit breaker

  1. 10 is grouped with load on circuit breaker #1. A new cable is routed between cabinets lBBDC and TP15-2 to accomplish this. Two fuses, one on each leg of circuit breaker #10, are added for proper penetration protection. Two additional fuses, one on each leg of the circuit breaker, are added for future use since no spare breakers are left on the bus. A loss of power relay (74) is added for monitoring DC power supply. 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 92-080)

DCP# lEC-3186 Pkg. 1 "Steam Generator Feed Pump High Discharge Pressure Trip" - The purpose of this change 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 that Feed Pump discharge pressure exceeds a 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 Steam Generator Feedpump (SGFP) circuitry is regarding a trip of the Main Feedwater Pumps starting the motor driven Auxiliary Feedwater Pump. This proposal does not alter this function in any way.

(SORC 92-082)

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

DCP# lEA-1027 Pkg. 1 "Permanent Documentation of Removal of 11SW928 From the system" - The purpose of this change is to permanently document the removal of 11SW928 from the system. Operators working drawings were updated by T-Mod 91-071. 11SW928 was a 3/4" diaphragm valve installed as a test connection during construction. MMIS has it listed as a drain valve, but it's doubtful is was used due to it's size and location on the 24" diaphragm header. During normal operation the valve was in the closed position. The valve is not used during any abnormal or emergency procedures. Removal of the valve has had minimal impact on station operations. The history of this valve is incomplete. 11SW928 was installed around 5/76, it was completely removed and a temporary repair done between then and the last refueling outage when the repair was discovered during an inspection of the large bore service area piping. It was thought to be a repair of the "thru wall hole" in the pipe, there was no indication of a valve. A code repair was done per deficiency report SMD-91-133. The missing valve was detected during the INPO audit of 07/29/91 to 08/09/91. The only consequences of this change are enhanced system reliability and reduced potential for leakage, neither of which affect the margin of safety.

(SORC 92-083)

B. Procedures and Revisions NC.NA-AP.ZZ-0069(Q) "Work Control Coordination", Rev. O - The purpose of this procedure is to describe the control process for coordination of work activities between implementing work organizations and the operating shift. This proposal does not involve any section of the Technical Specifications, therefore, the proposal cannot reduce the margin of safety as defined in the basis for any Technical Specification. {SORC 92-076)

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

c. Temporary Modifications TMR# 92-044 "Enhance Leaktightness of Air-Operator Diaphragms for Valve 2PR1, 1PR2, lPSl, 1PS3" - The air-operator for valve lPRl, 1PR2, lPSl, and 1PS3 will be temporarily modified by installing a backup o-ring. The o-ring will act as a redundant seal to the existing diaphragm. This sizing and material selection of the o-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 0-ring and diaphragm is anticipated based on similar material properties. The dimensions of the 0-ring are sized such that no interferences with the actuator movement will occur. Lab testing had been performed at elevated temperatures that have postulated to occur in the installed location. No failure modes resulted from the installation of the 0-ring at these temperatures. Degradation of 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 T-Mod. No adverse interaction between the 0-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.

(SORC 92-083)

D. SAR Change Notice (SCN)

SCN# 92-42 "Updating the SAR" - The purpose of this safety evaluation is to update the SAR to support the conclusions that the Salem Generating Station (SGS) control room will remain habitable during a postulated accidental release of any of the hazardous chemicals stored onsite or delivered to the SGS. The changes in the SAR do not reflect a reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 92-076)

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH:- - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (cont'd)

ITEM

SUMMARY

E. Technical Specification Interpretation (TSI)

TSI# 3

  1. DES-92-00256. The reduced MSSV flows were conservatively determined and are documented in NFU-92-424, attached. The Salem licensing basis was evaluated and it was determined that the reduced MSSV flows would not result in exceeding any safety limit provided that: 1.) At any power level above 55% RTP, all MSSVs are operable and 2.) At 55% RTP or below (including MODES 2 and 3) a maximum of 1 MSSV per line (up to a total of 4) may be inoperable. The reduced MSSV relief capacity does not reduce the margin of safety of any Technical Specification since it has been demonstrated that all safety limits are met.*

(SORC 92-081)

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 JULY 1992 SALEM UNIT NO. 1 During July the Unit was in a forced outage due to erosion/corrosion concerns with feedwater piping. This forced outage was entered on June 26, 1992, during the 10th Refueling Outage.

REFUELING INFORMATION DOCKET NO: 50-272 MONTH:. - JULY 1992 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1992 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 MONTH JULY 1992

1. Refueling information has changed from last month:

YES NO X

2. Scheduled date for next refueling: OCTOBER 2, 1993
3. Scheduled date for restart following refueling: DECEMBER 13, 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?:

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 193
b. In Spent Fuel Storage 656
8. Present licensed spent fuel storage capacity: 1170 Future spent fuel storage capacity: 1170
9. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: September 2001 8-l-7.R4
Retrieved from "https://kanterella.com/w/index.php?title=ML18096A888&oldid=2409130"