ML18096A391
| ML18096A391 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/30/1991 |
| From: | Fest J, Polizzi V, Shedlock M Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9112200227 | |
| Download: ML18096A391 (14) | |
Text
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Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station December 13, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 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 November 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 Sincerely yours,
~Wo4r
~~l Manager -
Salem Operations cc:
Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA Enclosures 8-1-7.R4 200006
__ Ihe_Enermd?_eooJe 9 :l l 2200227 9 :l l 1 ::::o PDR ADOCK 05000311 R
PDR 19046 95-2189 (10M) 12-89
I I '
~VERAGE DAILY UNIT POWER ~L Docket No.:
50-311 Unit Name:
Salem #2 Date:
12/10/91 Completed by:
Mark Shedlock Telephone:
339-2122 Month November 1991 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET)
(MWe-NET) 1 1076 17 0
2 1091 18 0
3 1090 19 0
4 1098 20 0
5 1088 21 0
6 1102 22 0
7 1088 23 0
8 1041 24 0
9 436 25 0
10 0
26 0
11 0
27 0
12 0
28 0
13 0
29 0
14 0
30 0
15 0
31 16 0
P. 8.1-7 R1
OPERATING DATA REPOR~
Docket No:
50-311 Date:
12/10/91 Completed by:
Mark Shedlock Telephone:
339-2122 Operating Status
- 1.
Unit Name Salem No. 2 Notes
- 2.
Reporting Period November 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.~~~N""'""'"'A..._~~~~~~~~~~~~~~~~~~~~~~~
- 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 720 203.3 0
203.3 0
689894.4 227880 214816 28.2 28.2 27.0 26.8 71.8 Year to Date Cumulative 8016 88825 7259.9 58616.1 0
0 7188.7 56898.7 0
0 24118252.8 130111721.8 7995121 59727048 7662334 56870145 89.7 64.1 89.7 64.1 86.4 57.9 85.7 57.4 7.3 22.5
- 24. Shutdowns scheduled over next 6 months (type, date and duration of. each)
We are presently in a maintenance and refueling outage.
25~ If shutdown at end of Report Period, Estimated Date of Startup:
April 15, 1992 8-1-7.R2
NO.
DATE 0081 11/08/91 0080 11/09/91 1
F:
Forced S:
Scheduled DURATION TYPE1 (HOURS)
REASON2 F
10.1 H
F 516.7 A
2 Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refueling D-Requlatory Restriction UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH NOVEMBER 1991 METHOD OF SHUTTING DOWN REACTOR 5
3 3
LICENSE EVENT REPORT #
Method:
1-Manual 2-Manual Scram SYSTEM CODE4 EB rF 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)
DOCKET NO. :-'5""0'--=3.:...,11;,,,....... __ _
UNIT NAME:
Salem #2 COMPONENT CODE5 DATE:
12/10/91 COMPLETED BY:
Mark Shedlock TELEPHONE:
339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE TRANF SOLAR MAGNETIC DISTURBANCES INSTRU TURBINE TRIP DEVICE FAILURE 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same Source
SkFETY RELATED MAINTENANCE MONTH: -
NOVEMBER 1991 DOCKET NO:
UNIT NAME:
50-311 SALEM 2 WO NO UNIT 901011145 2
901201307 2
910717233 2
910717238 2
911029133 2
911030200 2
911117115 2
~--
VALVE 22MS44 DATE:
COMPLETED BY:
TELEPHONE:
DECEMBER 10, 1991 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION FAILURE DESCRIPTION:
22MS44 GASKET FLANGE STEAM LEAK -
REPAIR VALVE 24SW15 FAILURE DESCRIPTION:
24SW15 IS LEAKING THROUGH -
OPEN AND INSPECT VALVE 2CH59 FAILURE DESCRIPTION:
2CH59 WILL NOT CLOSE -
REPLACE VALVE 22SW960 FAILURE DESCRIPTION:
REPLACE GASKET VALVE 22SW239 FAILURE DESCRIPTION:
VALVE 22SW239 LEAK IN WELD -
PERFORM UT READINGS 2R45B PLANT VENT RADIATION MONITOR FAILURE DESCRIPTION:
2R45B PLANT VENT RADIATION MONITOR FAILED LOW -
INVESTIGATE VALVE 21CA330 FAILURE DESCRIPTION:
21CA330 FAILED ITS LEAK RATE TEST TROUBLESHOOT
10CFR50.59 EVALUATIONS MO'l\\TTH: -
NOVEMBER 19 91 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 DECEMBER 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)
DCP # 2EC-3099 Pkg. 1 DCP # 2EC-3094 Pkg. 1 DCP # 2EC-3041 Pkg. 1 "Waste Liquid Disposal System Piping Reclassification" -
The purpose of this design change is to reclassify piping downstream of the 21, 22 and 23 BR-81 valves which are the 21, 22 and 23 eves Holdup Tank relief valves respectively.
The present piping is class 53D, nuclear III, non-safety related, seismic III.
The revised classification will be 53S, non-nuclear, non-safety related.
The reclassification of this piping is consistent with the present codes and regulatory guidelines governing radioactive waste management systems.
(SORe 91-112)
"Steam Generator Blowdown Maintenance Isolation Valves" -
This design change package replaces and relocates the existing steam generator blowdown maintenance isolation valves with new valves for better reliability and access.
The new valves will have reduced maintenance requirements.
The new location of the valves will improve ALARA considerations by eliminating the need to enter containment.
The replacement of pipe spools and valves will not impact the Technical Specifications.
The change does not alter the original design intent or the modes of operation or function for which the steam generator blowdown system is currently analyzed.
Therefore, the margin of safety basis for the Technical Specifications will not be affected.
(SORC 91-112)
"Service Water Pipe Upgrade to 6% Maly" - This design change will upgrade the 21 Service Water (SW) Header supply and return unit cooler piping from the connection at the 20" supply to the connection at the coil for each room cooler heat exchanger and equipment cooler at the ECCS pumps.
10CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM DCP # 2EC-3073 Pkg. 4 DCP # 2EC-3041 Pkg. 2 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 The temporary and permanent conditions of this DCP do not increase the probability or consequences of an accident or malfunction of equipment previously evaluated in the SAR.
The proposed pipe system modification meets the design requirements set forth in FSAR Sections 3.2, 3.6, 6.3 and 9.2.
Any risk due to a medium energy line break is enveloped by piping already located in the areas of installation.
(SORC 91-114)
"Unit 2 Main Steam Isolation Valve Limit Switch Modification" -
This design change will replace the Namco limit switches on each of the four Main Steam Isolation Valves (MSIV) with new limit switches of the same make and type; but with factory mounted quick disconnect receptacles.
In addition, the mounting configuration of the switches is being modified with new stainless steel standoffs and 1/4" CEM-FIL glass insulation to reduce heat conduction from the mounting brackets to the limit switches, which should result in lower temperatures to the switches.
The replacement of the limit switches does not change the facility as described in the FSAR since the new switches will be identical in function and type.
The bracket modification does not involve any functional change nor does it change the seismic characteristics of the of the MSIVs or brackets.
This proposal will not reduce the margin of safety as defined in the basis for any Technical Specification because it will not change the identity, function, or operation of the limit switches.
(SORC 91-117)
"Service Water Piping Upgrade to 6% Moly" - This design change will upgrade the 22 Service Water Header supply and return unit cooler piping from the connection at the 24" supply I 20" return to the connection at the coil for each room cooler heat exchanger and equipment cooler at the ECCS pumps.
10CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 The interim installation and test conditions do not impact the ability of the Service Water System to perform its intended safety function and will be in full compliance with the Technical Specifications.
A stress analysis has been performed to assure the temporary configuration that will exist during the outage is acceptable and can be declared operable to meet the Technical Specification requirement for one header operable during refueling.
The final configuration does not change the normal or emergency operation of the Service Water System with the exception of the cross connect capability.
The use of this cross connect will be administratively controlled to only those times during which one main header is out of service, increasing the available components.
This DCP does not affect the margins of safety defined by the Technical Specifications.
(SORC 91-119)
DCP #2EC-3084 Pkgs 1-3 "91.6% Undervoltage Relay Replacement" -
The proposed change will replace all nine 91.6%
Undervoltage relays that monitor 2A, 2B, and 2C Vital Busses for Undervoltage/Blackout conditions.
The actual change will involve removing the old UV relays from three vital bus switchgear cabinet doors and installing the three 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 each relay's wiring and to make testability more convenient.
All of the equipment has been evaluated for Seismic I, Class 1E use.
Chapter 15 of the SAR has been reviewed and there is no increase in the probability or consequences of an accident as described.
The new model 27N relay is Class lE and seismically approved and is considered better than an equivalent replacement.
The function of the new relay remains the same as that of the existing UV and time delay relays and does not affect any accident previously described.
(SORC 91-119)
10CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM B.
Temporary Modifications TMR # 91-098 TMR # 91-102 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 "23A Hotwell Conductivity Instrumentation" -
The 23A hotwell cation conductivity instrument tubing will be modified such that the supply line will be fed from the return root valve, and that the return line will be fed from the supply root valve.
This is being done to keep the instrument working while troubleshooting is being performed on the loop to correct a no-flow condition.
The hotwell cation conductivity instrument loop is non-nuclear, non-seismic, non-safety related and not important to safety.
The conductivity signal is not tied to any automatic control function.
There is no equipment important to safety located in the immediate vicinity of the temporary modification.
(SORC 91-111)
"22AR22 Blind Flange" - This TMOD provides documentation for a previously undocumented blind flange which was installed on valve 22AR22.
During an earlier system modification an extra vacuum pump was added to the system.
At that time 23 Vacuum Pump was dedicated to priming operation.
Valve 22AR22 was blind flanged to prevent cross flow between the priming and vacuum portions of the system.
No accident scenarios are affected by the inability to manually tie 23 Vacuum Pump into the air removal system since the system design does not rely on the ability to use the pump.
Therefore, its loss would not create the possibility of a malfunction or accident of a different type than previously evaluated in the SAR or increase the consequences of an accident previously evaluated in the SAR.
(SORC 91-117)
10-CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM TMR # 91-111 TMR # 91-109 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 "Use of a Temporary Pump to Drain 21 & 22 Service Water Nuclear Headers" - This TMOD provides for a temporary pump to be utilized for draining 21 & 22 Service Water Nuclear Headers and Containment Fan Coil Unit Headers during the 6th Refueling Outage of Unit 2.
By the addition of a drain pump this TMOD creates a different drainage path for Service Water drainage as described in Section 9.2 of the FSAR.
This is the only change to the facility and has no bearing on the accident analyses in Section 15 of the FSAR.
The water to be drained is to be analyzed to ensure that the criteria in Section 11.2 of the FSAR are met.
Therefore, there is no increase in the probability of an accident or malfunction of equipment.
Also, there is no increase in the consequences of an accident or malfunction of equipment.
( SORC 91-119)
"Use of a Temporary Mixed Bed Demineralizer" -
This TMOD involves interconnecting a temporary mixed bed demineralizer vessel and pump to the Unit 2 Primary Water Storage Tank (PWST).
This demineralizer will be operated in a recycle mode, taking water from and returning it to the PWST.
Neither the PWST nor the Radwaste Demineralizer system is important to safety.
The hose will not be located near any safety related equipment in the service pipe tunnel that can be negatively impacted by wetting.
All of the equipment, piping, etc. which could possibly be wetted by a hose rupture is located outdoors and is designed for outdoor installations.
Any inadvertent water spray that might occur would be minor as compared to weather conditions normally experienced by this equipment.
Therefore, the proposal does not increase the probability of occurrence of an occurrence of a malfunction to equipment important to safety previously evaluated in the SAR.
( SORC 91-119)
10CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM C.
Procedures and Revisions SC.RP-TI.ZZ-1130(Q)
VS2.RE-FR.ZZ-0003(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 "Laboratory Testing of Carbon Adsorbers in the Unit 2 auxiliary Building Ventilation System" -
This procedure change revises the existing acceptance criteria for laboratory testing of carbon adsorbers in the Unit 2 Auxiliary Building Ventilation System.
The margin of safety as defined in the Technical Specifications is not affected.
Sufficient margin over FSAR assumed charcoal efficiencies is maintained after selection of our new acceptance criteria.
The basis for Technical Specification 3.7.7 states that "ANSI N510-1975 and Generic Letter 83-13 should be used as procedural guidelines for surveillance testing".
The procedure acceptance criteria selected is in fact more conservative than Generic Letter 83-13 requires.
(SORC 91-113)
"Westingouse Refueling Procedure FP-PNJ-R6" -
This is a new refueling procedure for Salem Unit 2's sixth refueling.
This procedure details refueling equipment checkouts, full core unload, control rod drive shaft latching and unlatching, insert changeouts, core mapping, and irradiated specimen removal.
The proposed procedure gives direction for a full core unload instead of a fuel shuffle.
The procedure contains a prerequisite to ensure proper boron concentration for refueling as specified in Technical Specification 3/4.9.1, and the procedure uses existing plant equipment designed for removing fuel.
Therefore, the proposed procedure will not increase the probability of an uncontrolled boron dilution during refueling or a fuel handling accident previously evaluated in the UFSAR.
By following approved existing site procedures and processes, and maintaining the requirements of the Technical Specifications, the margin of safety for the performance of a full core unload will be bounded by the existing analyses for a core shuffle as described in UFSAR Section 9.
Therefore, the proposal does not reduce the margin of safety as defined in the basis for any Technical Specification.
(SORC 91-117)
10CFR50.59 EVALUATIONS MONTH: -
NOVEMBER 1991 (Cont'd)
ITEM II-3.3.6 Rev. 12 DOCKET I\\TO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 DECEMBER 10, 1991 J. FEST (609)339-2904 "Boron Concentration Control" - This revision allows the Demineralized Water Transfer System to be used as a source of water for boron concentration control.
The current procedure allows only the use of the Primary Water Storage Tank water to perform this evolution.
Since this change does not decrease the quality of water entering the RCS or alter or impact the operation of any safety related equipment, the change will not increase the possibility of an accident of a different type than any previously evaluated in the SAR.
Also, this change will not increase the probability of occurrence or the consequences of any accidents previously evaluated in the SAR.
( SORC 91-119)
SALEM UNIT NO. 2 SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 2 NOVEMBER 1991 The Unit began the period operating at full power and continued operating at full power until November 8, 1991, when power was reduced to 80% due to a Solar Magnetic Disturbance.
The Unit was restored to 100% power on November 9, 1991, at 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br />.
At 1121 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.265405e-4 months <br />, the same day, a unit turbine/reactor trip occurred due to a turbine hydraulic low oil pressure signal following completion of the final test of the scheduled monthly surveillance test, "Turbine Automatic Trip Mechanism Operational Test".
Operators were returning the equipment to normal when the trip occurred and an overspeed condition ensued and resulted in turbine blades penetrating the #22 low pressure turbine casing.
A hydrogen fire developed in the generator exciter area and ignited leaking lube oil.
The plant fire brigade responded alertly and extinguished the fire within 15 minutes.
A damage assessment is in progress with the assistance of Westinghouse and General Electric expertise.
The Unit is currently in Mode 6, defueled, and starting the sixth refueling outage ahead of the January 4, 1992 schedule due to the extensive turbine generator damage.
Ji j
REFUELING INFORMATION MONTH: -
NOVEMBER 1991 DOCKET NO:
UNIT NAME:
50-311 SALEM 2 DATE:
COMPLETED BY:
TELEPHONE:
DECEMBER 10, 1991 J. FEST (609)339-2904 MONTH NOVEMBER 1991
- 1.
Refueling information has changed from last month:
YES X
NO
- 2.
Scheduled date for next refueling:
NOVEMBER 11, 1991
- 3.
Scheduled date for restart following refueling:
APRIL 1, 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?:
December 1991
- 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 0
- b.
In Spent Fuel Storage 601
- 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:
March 2003 8-1-7.R4
- - Refueling outage dates may be revised due to turbine generator failure.