ML18096A352
| ML18096A352 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/31/1991 |
| From: | Fest J, Shedlock M Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9111220197 | |
| Download: ML18096A352 (13) | |
Text
.
e OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station November 14, 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 October 1991 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 Challenges to Safety Valves ly yours, cc:
Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4 The Energy People
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AVERAGE DAILY UNIT POWER LEVEL Completed by:
Mark Shedlock Month October 1991 Day Average Daily Power Level (MWe-NET) 1 1103 2
1097 3
1098 4
1112 5
1122 6
1103 7
1101 8
1096 9
1103 10 1109 11 1119 12 1108 13 1104 14 1115 15 1108 16 1120 P. 8.1-7 R1 Day Docket No.:
50-272 Unit Name:
Salem #1 Date:
11/10/91 Telephone:
339-2122 Average Daily Power Level (MWe-NET) 17 1108 18 1116 19 1105 20 1120 21 1110 22 1117 23 1110 24 1110 25 1120 26 1120 27 1102 28 1109 29 977 30 1100 31 1114
OPERATING DATA REPORT Docket No:
50-272 Date:
11/10/91 Completed by:
Mark Shedlock Telephone:
339-2122 Operating Status
- 1.
Unit Name Salem No. 1 Notes 2.
Reporting Period October 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 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 745 745 0
745 0
2534260.8 856920 823287 100 100 99.9 99.1 0
Year to Date 7296 5172.8 0
5016.4 0
16611914.4 5500150 5244952 68.8 68.8 65.0 64.5 5.4 Cumulative 125689 82136.3 0
79583.9 0
250674299.6 83212790 79213624 63.3 63.3 57.0 56.5
- 21. 6
- 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
Refueling Outage scheduled to start 4-4-92 and last 67 days.
- 25. If shutdown at end of Report Period, Estimated Date of Startup:
NA 8-1-7.R2
NO.
DATE 0061 10/28/91 1
2 F:
Forced S:
Scheduled DURATION TYPE1 (HOURS)
REASON 2 F
7.6 Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refuel ing D-Requlatory Restriction H
UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH OCTOBER 1991 METHOD OF SHUTTING DOWN REACTOR 5
3 LICENSE EVENT REPORT #
Method:
1-Manual 2-Manual Scram SYSTEM CODE4 EB 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.: ""'5""'0'°'"-=27"""'2""'
0 UN IT NAME : _s""'a"-'l'-"e'='m""'#=-:1;----
DATE:
11/10/91 COMPLETED BY:
Mark Shedlock TELEPHONE:
339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE TRAN SF SOLAR MAGNETIC DISTURBANCES 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same Source e
SAFETY'kELATED MAINTENANCE MON~H: -
OCTOBER 1991 DOCKET NO:
UNIT NAME:
50-272 SALEM 1 WO NO UNIT 901030198 1
910829103 1
910916128 1
910918174 1
911002182 1
911014245 1
911023115 1
DATE:
COMPLETED BY:
TELEPHONE:
NOVEMBER 10, 1991 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION VALVE 12SW102 FAILURE DESCRIPTION:
12SW102 LEAKS THROUGH -
INVESTIGATE SERVICE WATER SPOOL 1SW1014 FAILURE DESCRIPTION:
SW SPOOL 1SW1014 HAS A THROUGH WALL LEAK -
REPAIR PLANT VENT FLOW RECORDER 1FA8605 FAILURE DESCRIPTION:
1FA8605/PLANT VENT FLOW RECORDER ERRATIC -
INVESTIGATE 1S27C FAILURE DESCRIPTION:
THROUGH WALL LEAK IN THE 4" SERVICE PIPING -
REPAIR VALVE 1S.S33 FAILURE DESCRIPTION:
VALVE 1SS33 NOT MAKING CLOSED LIMIT -
INVESTIGATE RMS CHANNEL 1Rl8 FAILURE DESCRIPTION:
RMS CHANNEL 1R18 SPIKING -
TROUBLESHOOT 12 COMPONENT COOLING HEAT EXCHANGER FAILURE DESCRIPTION:
12 CCHX TEMPERATURE CONTROL FAULTY -
TROUBLESHOOT
10CFR50:59 EVALUATIONS MONrH: -
OCTOBER 1991 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-272 SALEM 1 NOVEMBER 10, 1991 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 # lEC-3046 Pkgs 1-3 "Boric Acid Transfer Pumps" -
These modifications propose to install a larger diameter shaft in the Boric Acid Transfer (BAT) pump.
A new frame adapter, frame foot, power end, stuffing box, and mechanical seal will be used to accommodate the new shaft.
The new stuffing box is designed with a tapered 'bore allowing increased circulation in the seal cavity which provides improved seal cooling.
Being that all hydraulic characteristics, manual and automatic controls, and pump to pipe connections will remain unchanged and overall pump integrity is improved with the new shaft, there is no unreviewed safety concerns pertaining to the completed pump upgrade.
(SORC 91-103)
DCP # lEC-3109 Pkg. 1 "91.6% Undervoltage Relays" -
The proposed change will replace all 9 91.6% undervoltage relays that monitor lA, 1B, and lC Vital Busses for undervoltage/blackout conditions.
Each vital bus uses three relays (i.e., 27-lA/l, 27-1A/2, and 27-lA/3 for "A" Bus) that monitors voltage between each phase.
The actual change will involve removing the old UV relays from the three vital bus switchgear cabinet doors and installing the 3 new UV time delay relays within the spare cubicle of each vital bus.
The existing time delay relays (TD-5) will be left in place and marked as spares.
In addition, a terminal board and test switch will be mounted adjacent to each relay in order to terminate the relay's wiring and make testing more convenient.
The new model relays are Class lE and seismically approved and are considered to be better than an equivalent replacement.
The function of the new relays remains the same as that of the existing undervoltage and time delay relays and does not affect any accident previously described.
( SORC 91-106)
10CFR50:59 EVALUATIONS MONYH:
~ OCTOBER 1991 (Cont'd)
ITEM DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-272 SALEM 1 NOVEMBER 10, 1991 J. FEST (609)339-2904 DCP # 5EC-3024 Pkg. 1 "Dewatering Wells" -
The proposed change will decommission several dewatering or observation wells on Artificial Island.
Additional dewatering/observation wells that may be found during the field installation or later may also be decommissioned using the procedures contained in this DCP.
This change will include the sealing and abandonment of wells and removal of cement pylons and steel protective casings from around the wells.
This change is an integral part of implementation requirements of the NJDEP Groundwater Diversion Permit 2216P. *The wells proposed for decommissioning have no function relative to operation or safety of the plant.
This is based on review of Section 2.4.13 of the UFSAR and on the fact that these wells are not production wells.
( SORC 91-108)
DCP # lEC-3102 Pkg. 1 "Ammonia & Hydrazine Injection" -
The purpose of this design change is to provide a permanent addition path for injection of ammonia and hydrazine into the Condensate Polishing discharge line.
The existing ammonia path, which runs from valve 1CF103 to a common injection line for the two chemicals, (upstream of 1CF87) will be modified to connect to valve 1CN127.
The tie in to the common injection path will be capped.
This change will remove and supersede temporary modification STD 89-071.
The additional chemical injection path will enhance the ability of the Chemistry Department to maintain proper chemistry control while the Condensate Polishing system is in service, thereby improving the reliability and integrity of the Feed & Condensate system chemistry.
There is no change to the function of any system and the change is to a non-safety related system.
Therefore, there is no possibility of an accident or malfunction of a different type than that evaluated in the SAR being created.
( SORC 91-110)
10CFR5~.59 EVALUATIONS MO~TH: -
OCTOBER 1991 (Cont'd)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-272 SALEM 1 NOVEMBER 10, 1991 J. FEST (609)339-2904
~-------------------------------------------------
ITEM
SUMMARY
B.
Procedures and Revisions (Proc)
NC.NA-AP.ZZ-0037(Q)
NC.NA-AP.ZZ-0005(Q)
VS1.RE-FR.ZZ-0002(Q)
"Environmental Control" -
Rev. 1 to this procedure adds specific measures for the rescue of impinged sea turtles.
This change implements the changes to the Salem and Hope Creek Environmental Protection Plans via Amendment 129/108 and 43 respectively.
This procedure revision does not involve a change to the facility or alter the design, function, or method of performing the function of any structure, system or component.
The proposed change does not create the possibility of an accident or malfunction of a different type than any previously evaluated in the UFSARs because the change relates to the observance and rescue of sea turtles which are impinged.
The plants are already designed to withstand sea turtle impingement.
(SORC 91-105)
"Station Operating Practices" -
Rev. 1 to this procedure incorporate changes made in response to the Salem SERT report dated January 24, 1991, provides additional information to the notification requirements, provides Salem communications examples, deletes the Fire Brigade Technical Liason role of the SNSS in accordance with NC.NA-AP.ZZ-0025(Q), and makes editorial changes.
Since the changes in this procedure are in compliance with the SAR, the proposal does not increase the probability or consequences of an accident previously evaluated in the SAR.
Since the changes in this procedure are in compliance with the SAR, the proposal does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.
(SORC 91-106)
"Fuel Recon Procedure" - This is a Westinghouse procedure used to guide the fuel repair operations on Salem Unit 1.
10CFR50~59 EVALUATIONS MON.TH: -
OCTOBER 1991 (Cont'd)
ITEM NC.NA-AP.ZZ-0023(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-272 SALEM 1 NOVEMBER 10, 1991 J. FEST (609)339-2904 This procedure contains instructions for the Multipurpose Fuel Repair System (MFRS) equipment assembly, checkout, installation and removal; bottom nozzle removal and reinstallation; fuel assembly reconstitution (the replacement of a defective fuel rod with a stainless steel rod);
fuel assembly reassembly (the transfer of all fuel rods from a defective skeleton to a new skeleton); and visual inspection of each failed fuel rod.
The procedure will not reduce the margin of safety as defined in the basis for the Technical Specifications since no T/S Limiting Condition for Operation will be violated.
Radiation monitors 1R5 and 1R9 will be operable as required along with the fuel handling building ventilation system in accordance with T/S 3.3.3 and 3.9.12 respectively.
No loads in excess of 2200 lbs will be carried over the spent fuel pool in accordance with T/S 3.9.7, and normal level in the spent fuel pool will be maintained in accordance with T/S 3.9.11.
- Also, the specific activity of the reactor coolant system will be adhered to for all future core reloads, ensuring that the offsite dose will be within the 10CFR100 limits in the event of a steam generator tube rupture.
Therefore, the margin of safety is not reduce by this procedure.
(SORC 91-106)
"Scaffolding and Transient Loads Control" -
Rev.
0 of this procedure describes the controls established for erecting and storing scaffo.lding in both nonsafety and safety related areas, including periodic inspections and walkdowns prior to startup and operation.
- Also, requirements and restraint guidelines for the handling and storage of transient loads are provided.
The issuance of this procedure does not reduce the margin of safety as defined in the basis for any Technical Specification.
(SORC 91-109)
10CFR50:59 EVALUATIONS MONTH: -
OCTOBER 1991 (Cont'd)
ITEM Sl-SO.WG-008,9,10,11 DOCKET NO:
UNIT NAME:
DATE:
CbMPLETED BY:
TELEPHONE:
SUMMARY
50-272 SALEM 1 NOVEMBER 10, 1991 J. FEST (609)339-2904 "Discharge of the Gas Decay Tanks to the Plant Vent" Rev. 3 of these procedures involves allowing the use of jumpers to defeat/block the closing signals to the WG41 valves from the RMS channel 1R41C when it is inoperable.
The installation of the jumper will permit the opening of the WG41 Gaseous Waste Discharge Isolation Valve.
The proposed change to the facility does not reduce the margin of safety as defined in the basis for any Technical Specifications.
Technical Specifications for Radioactive Gaseous Effluent Monitoring Instrumentation 3/4.3.3.9 has already addressed the Limiting Condition for Operations via Action Statement 31, when R41C is unavailable.
The proposed change allows the Action Statement to be used.
The installation of the jumpers will not disable any other parts of the WG41 control circuitry.
(SORC 91-110)
SALEM UNIT NO. 1 SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 1 OCTOBER 1991
.The Unit began the period operating at full power and continued to operate at full power until October 28, 1991, when load was reduced to 76% due to a Solar Magnetic Disturbance.
The Unit was restored to 100% power on October 29, 1991, and continued to operate at
- essentially full power throughout the remainder of the period.
REFUELING INFORMATION MO~TH: -
OCTOBER 1991 DOCKET NO:
50-272 UNIT NAME:
SALEM 1 DATE:
COMPLETED BY:
TELEPHONE:
NOVEMBER 10, 1991 J. FEST (609)339-2904 MONTH OCTOBER 1991
- 1.
Refueling information has changed from last month:
YES NO X
- 2.
Scheduled date for next refueling:
APRIL 4, 1992
- 3.
Scheduled date for restart following refueling:
JUNE 9, 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
CHALLENGES TO PORVs OR SAFETY VALVES Salem Unit 1 In accordance with the requirements of Salem Generating Station Unit 1 Technical Specifications Section 6.9.1.6, the following recent challenges to PORVs or Safety Valves are being reported:
09/18/91 -
On September 18, 1991, with Salem Unit 1 in Mode 3, during the performance of RTD cross calibrations, Steam Generator Power Operated Relief Valve 14MS10 went full open.
The valve was placed in manual and closed along with its blocking valve, 14MS9.
Subsequently, Main Steam Safety Valve 14MS15 lifted and began to chatter. Reactor Coolant temperature was then lowered and the valve reseated.
At the time of the event all four Tave channels were out of service for time response testing and Tave was being monitored using the P-250 computer.
Prior to 14MS10 lifting, the operators had been in the process of raising Tave from 539°F to 547°F to assist in the RTD cross calibrations.
The highest steam pressure observed was
-1040 psig.
The account of this event was inadvertently omitted from the Unit 1 September 1991, monthly report.