ML18096A268
| ML18096A268 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/31/1991 |
| From: | Fest J, Shedlock M Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9109200259 | |
| Download: ML18096A268 (14) | |
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Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station September 13, 1991 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 August 1991 are ~eing 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 Since ly yours,
~/ffd-eneral Manager -
Salem Operations RH:pc cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4 95-2189 (10M) 12-89
/
A~GE DAILY UNIT POWER LEV~
Docket No.: 50-272 Unit Name: Salem #1 Date: 9/10/91 Completed by: Mark Shedlock Telephone: 339-2122 Month August 1991 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 1069 17 1078 2 1083 18 1055 3 1056 19 1084 4 1062 20 1078 5 1076 21 1059 6 1061 22 1064 7 1088 23 1047 8 1071 24 533 9 1077 25 521 10 1071 26 524 11 1077 27 534 12 1062 28 701 13 1022 29 1024 14 347 30 1088 15 572 31 1062 16 942 P. 8.1-7 Rl
e OPERATING DATA REPORT e Docket No: 50-272 Date: 9/10/91 Completed by: Mark Shedlock Telephone: 339-2122 Operating Status
- 1. Unit Name Salem No. 1 Notes
- 2. Reporting Period August 1991
- 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 Reasori.~~N___A~~~~~~~~~~~~~~~~~~~~~~~~-
- 9. Power Level to Which Restricted, if any (Net MWe) N/A
- 10. Reasons for Restrictions, if any NA This Month Year to Date Cumulative
- 11. Hours in Reporting Period 744 5831 124224
- 12. No. of Hrs. Rx. was Critical 744 3927.9 80891. 4
- 13. Reactor Reserve Shutdown Hrs. 0 0 0
- 14. Hours Generator on-Line 744 3812.5 78380.0
- 15. Unit Reserve Shutdown Hours 0 0 0
- 16. Gross Thermal Energy Generated (MWH) 2279188.8 12541836.0 246604221. 2
- 17. Gross Elec. Energy Generated (MWH) 733730 4134600 81847240
- 18. Net Elec. Energy Gen. (MWH) 700511 3940473 77909145 19~ Unit Service Factor 100 65.4 63.1
- 20. Unit Availability Factor 100 65.4 63.1
- 21. Unit Capacity Factor (using MDC Net) 85.1 61. 6 56.7
- 22. Unit Capacity Factor (using DER Net) 84.4 60.6 56.2
- 23. Unit Forced Outage Rate 0 4.9 21. 8
- 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-l-7.R2
UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH AUGUST 1991 DOCKET NO.: -'5"""'0'---=-27'""'2.....,__ __
UNIT NAME: Salem #1 DATE: 9/10/91 COMPLETED BY: Mark Shedlock TELEPHONE: 339-2122 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE 1 (HOURS) REASON 2 REACTOR REPORT # CODE 4 CODE 6 . TO PREVENT RECURRENCE 0049 8/14/91 F 37.1 A 5 ------ IA INSTRU REACTOR COOLANT TEMP SENSOR 0054 8/23/91 F 112.4 B 5 ------ CH INSTRU #12 STEAM GEN. FEED PUMP 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) Fi le D-Requlatory Restriction . 4-Continuation of (NUREG-0161)
E-Operator Training & License Examination Previous Outage F-Administrative 5-Load Reduction G-Operational Error (Expl~in) 9-0ther H-Other (Explain)
I '
MONTH: - AUGUST 1991 SAFETY.RELATED MAINTENANCE DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 WO NO UNIT EQUIPMENT IDENTIFICATION 910612120 1 #11 BORIC ACID TRANSFER PUMP FAILURE DESCRIPTION: BORIC ACID TRANSFER PUMP IS LEAKING OIL - INVESTIGATE 910806098 1 NUCLEAR INSTRUMENT 1N41 FAILURE DESCRIPTION: INSTALL NEW NI DETECTOR CURRENT PER FLUX MAP 910806100 1 NUCLEAR INSTRUMENT 1N42 FAILURE DESCRIPTION: INSTALL NEW NI DETECTOR CURRENT PER FLUX MAP 910806102 1 NUCLEAR INSTRUMENT 1N43 FAILURE DESCRIPTION: INSTALL NEW NI DETECTOR CURRENT PER FLUX MAP 910806104 1 NUCLEAR INSTRUMENT 1N44 FAILURE DESCRIPTION: INSTALL NEW NI DETECTOR CURRENT PER FLUX MAP 910813111 1 12 SERVICE WATER STRAINER FAILURE DESCRIPTION: 12 SW STRAINER HAS A HIGH DP -
TROUBLESHOOT 910814077 1 11 H2 MONITOR FAILURE DESCRIPTION: 11 H2 MONITOR INPUT ERRORS -
TROUBLESHOOT 910815146 1 lA SAFEGUARDS EQUIPMENT CABINET FAILURE DESCRIPTION: lA SEC FAILED AND CAUSED A PARTIAL ACTUATION - INVESTIGATE 910822085 1 RADIATION MONITOR 1R45 FAILURE DESCRIPTION: 1R45 MONITOR CONTROL TERMINAL ALARMING - INVESTIGATE 910823118 1 VALVE 12GB4 FAILURE DESCRIPTION: 12GB4 WILL NOT OPEN FROM CONTROL ROOM - REWORK
10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - AUGUST 1991 UNIT NAME: SALEM 1 DATE: SEPTEMBER 10, 1991 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 # lEC-3047 Pkg. 3 "Service Water Crossover Piping" - Package 3 proposes adding a 16" flanget and a 16" blind flange to the 30" crossover pipe between bay #2 to bay #4 between valves 21SW17 and 22SW17.
This is package 3 of three packages designed to add a separate test line td allow full flow testing of the the Service Water pumps. The addition of the flanget and the blind flange is acceptable to the pipe stress supports. All loads have been determined to be acceptable with no modifications required. This will have no impact on the operation of the system. The new carbon steel flanget is fully compatible with the system design requirements for existirig pipe and will be lined consistent with current practices for replacing carbon steel pipe in the service water system.
(SORC 91-060)
DCP # lEC-3047 Pkg. 2 "Service Water Crossover Piping" - Package 3 proposes adding a Service Water pump full flow test line to the Service Water Intake Structure. The purpose of this line is to allow Operations to reliably perform Service Water pump Surveillance Testing. The added Nuclear class 3 pipe material is 6% moly stainless steel and is compatible for use in the Service Water System. The added Nuclear Class 3 safety related double isolation butterfly valves are 316 stainless steel and are fully compatible for use in the Service Water System. The double isolation valves are normally closed and provide appropriate isolation to assure the pressure boundary integrity of the safety related portions of the Service Water System. The line is designed and installed to Seismic Class 1 to assure operability under all required accident conditions.
10CFR50.59 EVALUATIONS MONTH: - AUGUST 1991
- DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
During surveillance testing of one Unit, the other Unit will be unaffected and isolated by the double isolation valves.
(SORC 91-069)
DCP # lSC-2269 Pkg. 1 "Salem Electrical Distribution Upgrade" - This package consists of construction of foundations, underground work fence and roadway for the expansion of the 500KV Switchyard, installation of piles for Power Transformers T3 and T4 and bus support foundations. Since this modification does not functionally interface with and safety or Technical Specification related systems, components or structures, the margin of safety as defined in the basis for the Technical Specifications is not affected.
(SORC 91-062)
DCP # lEC-3061 Pkg. 1 "Re-Adjust 1R19A,B,C&D Setpoint Calculations" -
The proposed change involves recalculating the alarm and warning setpoints for the 1R19A,B,C&D Radiation Monitors. This was necessary because the dilution rate, Circulating Water (CW) flow rate was conservatively derated from the value used in the original calculation. This was based on information provided by recent testing. The aging of the pumps is given as the reason for the decreased flow rates. The proposed setpoint changes are more conservative than the present setpoints. Additionally, the criteria used to establish the new operating setpoints are in accordance with the standards established in the Salem SAR. ( SORC 91-063)
DCP # 5EC-3024 Pkg. 2 "Monitoring Well Decommissioning, Replacement, and Upgrade" - This package provides for decommissioning Well Nos. OW-A/C/D/H, Replacement Well OW-J to replace OW-A and upgrade Well Nos. OW-I/G/6. The wells being closed were installed to monitor the impact of groundwater of the operations of Hope Creek and Salem Generating Stations. During the consolidation of diversion permits, coinciding and in accordance with the renewal of the current Salem and Hope Creek Water Allocation Permits.
10CFR50.59 EVALUATIONS MONTH: - AUGUST 1991
- DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
These wells are no longer required, require modification, or are being abandoned due to questionable integrity. All modifications including well abandonment and sealing, well installation, and well upgrading will b~
implemented in accordance with the NJDEP rules, specific water diversion permit requirements for this DCP project and regulations set out in N.J.A.C.7:9-9et seq. (SORC 91-068)
DCP # lEC-3046 Pkg. 1 "BAT Flow Measurement Device" - This modi£ication involves the installation of a flow measurement device for Boric Acid Tank (BAT) pump surveillance testing. This modification does not reduce the margin of safety as defined in the basis for.any Technical Specification.
By installing this device, the requirements for surveillance testing in Section 4.0.5 will be met. (SORC 91-072)
DCP # lEA-1011 "As-Built - Ultrasonic Sink Removal" - This DCP is proposed to reflect as-built conditions in associated drawings and documents to show the elimination of the ultrasonic sink in the Decon.
Room, Elevation 100' of the Auxiliary Building.
Deletion of the ultrasonic sink does not reduce the margin of safety, but actually increases the availability of demineralized water and also creates more available storage space within the Waste Hold Up tanks. (SORC 91-073)
B. Temporary Modification Requests (TMRs)
TMR # 91-049 "Pneumatic Jumper" - The purpose of this modification is to provide remote standby diesel driven air compressors to act as a backup air supply while two of the station air compressors are repaired. The remote air compressors are to be installed to valve 12SA909 "Station Air Receiving Manifold". The remote compressors will comply with the design basis listed in SAR Section 9.3.1.1 and provide the capacities required of station air compressors in Section 9.3.1.2.1.
10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - AUGUST 1991 UNIT NAME: SALEM 1 DATE: SEPTEMBER 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
The three remote compressors will be connected at the 12SA909 valve manifold. This connection allows the air to discharge into two independent air headers as described in Section 9.3.1.2.1.
The control air system is not affected by adding the remote compressors. The compressed/station air receiving station is not safety related or required for safe shutdown. *
(SORC 91-064)
C. Procedures and Revisions (Proc)
Proc # SP7 "Personnel Access Control" Rev. 4 - This procedure is changed to reflect the use of a new visitor processing procedure and related forms.
Also, a new section, Photobadge Management is added to streamline and reduce the margin for error in the process. This change has no affect upon or connection with plant systems. This change is designed to enhance cOmmunications between the Badging Group and Security to ensure the accuracy of all changes, and, therefore, to reduce the chances of an error contributing to unauthorized access. (SORC 91-063)
Proc # SP8 "Vehicle Access Control" Rev. 5 - This procedure is changed in response to an NRC concern documented in Combined Inspection Report 50-272/91-1, 50-311/91-10 & 50-354/91-07 that wording in the procedure does not reflect current practices with regard to Temporarily Designated Licensee Vehicles (TDLVs). The change distinguishes a TDLV from a designated Licensee Vehicle in that a TDLV is permitted more frequent protected area entries and exits and its need for access revalidated every 31 days. (SORC 91-063)
Proc # SP12 "Security System Testing" Rev. 6 - This procedure is revised to eliminate retention of redundant forms for documentation of Security System testing. The forms in question have been renamed as worksheets and are used in the field to keep track of individual test components.
10CFR5D.59 EVALUATIONS MONTH: - AUGUST 1991
- DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
The fact that an alarm zone passed the overall test is documented on separate forms which are retained. The procedure change does not affect any plant operating system. The change does not affect the testing procedure which has been .
validated by the NRC Regulatory Effectiveness Review. (SORC 91-063)
Proc # SP15 "Safeguards Event Reporting" Rev. 7 - This procedure has been revised primarily to incorporate new guidance provided in NRC Generic Letter 91-03, Reporting of Safeguards Events.
The procedure change involves reporting security incidents to the NRC which may also be initiating events for implementation of the Emergency Plan. Nuclear plant equipment or operating procedures are not affected.
(SORC 91-063)
- Sl.RE-RA.ZZ-OOlO(Q) Guidelines for Misaligned/Dropped Control Rod Recovery" Rev. 0 - This procedure provides guidelines for the operation of the reactor during the recovery of a dropped or misaligned control rod. The procedure was written to satisfy the requirements of INPO SOER 84-02R01.
This procedure will not create the possibility of an accident of a different type than any previously analyzed in the SAR. The UFSAR and the Technical Specifications allow for the recovery of a misaligned or dropped control rod. The guidelines in the procedure were established to ensure that certain acceptance criteria are met during and after the recovery.
These criteria are within the UFSAR and Technical Specifications bounds. (SORC 91-075)
D. Safety Evaluations (SE)
SE # S-C-CBV-NSE-0797 "Containment Fan Coil Unit (CFCU) and Containment Spray (CS) Response Time Testing" The purpose of this evaluation is to address a discrepancy noted between the CFCU and CS response time testing and the response time assumed in the safety analyses.
I 10CFR5.0. 59 EVALUATIONS MONTH: - AUGUST 1991
- DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
The discrepancies between the Technical Specifications and accident analyses for the CFCUs and CS response times requires a change to the SAR. Specifically, reanalysis of containment response to limiting LOCA/MSLB cases using an increased CFCU respqnse time constitutes a change to the facility.
Reanalysis of the limiting cases for LOCA/MSLB using the increased bounding values for CFCU response time, demonstrate that the containment pressure and temperature response would remain within acceptable limits. The increased CFCU response time has no impact on postulated LOCA/MSLB radiological consequences, because the offsite dose calculations do not take credit for fission product removal by the CFCUs.
(SORC 91-065)
SE # HEBA "HEBA Barrier Penetration" - The purpose of this evaluation is to address Penetration Seal Work Release 1382 which has been issued to run temporary cable(s) through penetration 5-15511-012 located between the mechanical penetration area at Elevation 100' and the north south aisle. If an HELB occurred in the mechanical penetration area during the time the barrier is breached, it is possible for steam to propagate through the penetration into the auxiliary building at elevation 100'. A Probability Risk Assessment (PRA) has determined that the increased likelihood of core damage from this postulated accident would be 2.9 E-9/day or a total increase of 1.2 E-8 for this activity. This probability is judged to be negligible, and therefore, the proposed breach of the barrier is acceptable. (SORC 91-066)
SE # ROOM 145 "UFSAR Change Notice - Records Storage Room 145"
- Records Storage Room 145 in the Nuclear Department Administration Building does not fully comply with all of the NRC guidelines and committed to in the Hope Creek and Salem Updated Final Safety Analysis Reports (UFSARs).
-- - - -------------------------------------~--
10CFR5-0.59 EVALUATIONS MONTH: - AUGUST 1991
- DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
50-272 SALEM 1 SEPTEMBER 10, 1991 J. FEST TELEPHONE: (609)339-2904 (Cont'd)
ITEM
SUMMARY
The ability of the facilities to achieve and maintain safe shutdown in the event of a fire is not directly related to the records storage facility in the Nuclear Department Administrative Building. The technical justification discussed in the evaluation provides sufficient assurance that important records will be adequately protected until they are duplicated on microfilm and stored in separate locations. (SORC 91-081)
SE # S-C-WL-MSE-0799 "Revision to the PCP and UFSAR - Radwaste Processing" - This SE documents the e.ffect of revising the process control program (PCP) and the sections of the UFSAR to reflect the current radioactive waste processing practices at Salem Generating Station. Technical Specification Section 3/4.11 was reviewed and it was determined that the use of a portable system in lieu of the waste evaporators and cement stabilized drumming system has no impact. The reason for this is that the procedure limits for the actual liquid release will remain the same, and also, the solid waste categorization and classification will remain the same.
(SORC 91-073)
- SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
- UNIT 1 AUGUST 1991 SALEM UNIT NO. 1 The Unit began the period operating at full power and continued to operate at essentially full power until August 14, 1991, when power was reduced to perform maintenance on Reactor Coolant System temperature devices. On August 15, 1991, the Unit began increasing power following replacement of #13 Reactor Coolant Loop resistance temperature detector. Response time testing of the RTD was completed satisfactorily at the 53% power plateau and power was increasing toward 100% when lA Safeguards Equipment Cabinet (SEC) failed at 98%.
A power decrease was initiated to comply with the Technical Specification Action Statement (TSAS). The SEC was repaired and the Unit returned to 100% power on August 16, 1991. The Unit continued to operate at full power until August 23, 1991, when load was reduced for scheduled repairs to #12 Steam Generator Feedwater Pump (SGFP) latch/trip mechanism. The repairs were completed and the Unit was returned to 100% power on August 28, 1991. The Unit continued to operate at full power throughout the remainder of the period.
REFUEkING INFORMATION DOCKET NO: 50-272 MONTH: - AUGUST 1991 UNIT NAME: SALEM 1 DATE: SEPTEMBER 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 MONTH AUGUST 1991
- 1. Refueling information has changed from last month:
YES NO X
- 2. Scheduled date for next refueling: APRIL 18, 1992
- 3. Scheduled date for restart following refueling: JUNE 12, 1992
- 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 588
- 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 8-1-7.R4