ML18096A513
| ML18096A513 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/31/1992 |
| From: | Shedlock M, Vondra C Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9202200098 | |
| Download: ML18096A513 (12) | |
Text
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CPS~G e Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station February 13, 1992 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 January 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
- yours, 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 Energy People
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9202200098 920131 PDR ADOCK 05000311 R
PDR 95-2189 (10M) 12-89
8.-1-7. R2 AVERAGE DAILY UNIT POWER LEVEL Docket No.:
50-311 Unit Name:
Salem #2 Date:
02/10/92 Completed by:
Mark Shedlock Telephone:
339-2122 Month JANUARY 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
10 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 Rl
P. 8. 1-7 R;t OPERATING DATA REPORT Docket No:
Date:
Completed by:
Mark Shedlock Telephone:
Operating Status
- 1.
Unit Name Salem No. 2 Notes
- 2.
Reporting Period January 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)
Report, Give Reason NA
- 9.
Power Level to Which Restricted, if any (Net MWe)
- 10. Reasons for Restrictions, if any
- 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)
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 NA This Month Year to Date 744 744 0
0 0
0 0
0 0
0 0
0 0
0
-1819
-1819 0
0 0
0 0
0 0
0 100 100 50-311 2/10/92 339-2122 since Last N/A Cumulative 90313 58616.1 0
56898.8 0
130111721.8 59727048 56866466 63.0 63.0 56.9 56.5 23.4
- 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 10,1992
NO.
DATE 0080 01/01/92
. 0001 01/04/92 1
2 F:
Forced S:
Scheduled DURATION TYPE1 (HOURS)
REASON2 F
n.o s
672.0 Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refueling D-Requlatory Restriction A
c UNIT SHUTDOYN AND POYER REDUCTIONS REPORT MONTH JANUARY 1992 METHOD OF SHUTTING DOYN REACTOR 4
4 3
LICENSE EVENT REPORT #
Method:
1-Manual 2-Manual Scram SYSTEM CODE4 IF RC 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" 02/10/92 Mark Shedlock 339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE INSTRU TURBINE TRIP DEVICE FAILURE FUELXX NUCLEAR NORMAL REFUELING 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same Source
SAFETY RELATED MAINTENANCE MONTH: -
JANUARY 1992 DOCKET NO:
UNIT NAME:
50-311 SALEM 2 WO NO UNIT 910123062 2
910318105 2
910503139 2
911125131 2
920107079 2
920118060 2
DATE:
COMPLETED BY:
TELEPHONE:
FEBRUARY 10, 1992 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION 2B DIESEL GENERATOR FAILURE DESCRIPTION:
ENGINE AIR MANIFOLD ERRATIC PRESSURE -
INVESTIGATE VALVE 2FP147 FAILURE DESCRIPTION:
REPLACE VALVE 2FP147 VALVE 23SW39 FAILURE DESCRIPTION:
VALVE LEAKS BY -
OPEN AND INSPECT VALVE 2FP148 FAILURE DESCRIPTION:
VALVE LEAKS BY -
OPEN AND INSPECT 22 CONTAINMENT HYDROGEN RECORDER FAILURE DESCRIPTION:
NO ADVANCE ON RECORDER -
INVESTIGATE & REPAIR FIT 111 FAILURE DESCRIPTION:
FIT 111 DOES NOT INDICATE FLOW -
INVESTIGATE
10CFR50.59 EVALUATIONS MONTH: -
JANUARY 1992 *
~--~~--~~-
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 FEBRUARY 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-3100 Pkg. 1 Narrow Range RTD Assembly Replacement" - This DCP will replace the Reactor Coolant System (RCS) narrow range resistance temperature detectors (RTDs) with new Weed Instrument RTDs which are the same model as the existing RTDs, however, the manufacturing process has improved the insulation resistance and the RTD is supplied with an EGS model bayonet style "quick disconnect".
The new RTDs will be provided with a butt splice in the termination head in lieu of a terminal block.
These changes will have no affect on the function of the RTDs.
The differences will result in more reliable RTDs and improved connection from the RTDs to the process cabinet.
Therefore, this DCP does not create the possibility of an accident or malfunction of a different type than previously evaluated.in the SAR.
Also, this DCP does not increase the probability or consequences of an accident previously evaluated in the SAR.
(SORC 92-009)
DCP# 2EC-3097 Pkgs 1-3 "2A, 2B, 2C Diesel Generator Modifications" Rev.
1 -
The purpose of this design change is to provide two temperature indicators each for the 2A, 2B and 2C Diesel Generators (D/Gs).
One of the temperature indicators is to monitor winding and bearing temperatures and the other replaces existing engine pyrometers.
This DCP also replaces the existing obsolete tachometers with Dynalco E.D.G speed switches to reduce D/G unavailability due to lack of spare parts.
The prelube oil pump motor breakers and thermal overload heaters will also be replaced.
The safety functions, control and operation of the D/Gs remain unchanged.
The replacement components are identical in function and performance and do not affect the UFSAR Chapters 3 and 15 analyses.
Therefore, the probability of occurrence or consequences of an accident previously evaluated in the SAR is not increased.
(SORC 92-009)
10CFR50.~9 EVALUATIONS MONTH: -
JANUARY 1992 (Cont'd)
ITEM DCP# 2EA-1013 DCP# 2SC-2267 Pkg. 1 DCP# 2EC-3102 Pkg. 1 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 FEBRUARY 10, 1992 J. FEST (609)339-2904 "Diesel Generator Load Calculations" - This DCP updates documents impacted by the revised Salem Diesel Generator (D/G) Load Calculation (ES-9.002-0) and confirms the adequacy of the D/Gs.
This calculation verifies that the maximum load demand would remain within the manufacturer's ratings which establish the margin of safety for DIG capability.
The ability of the D/Gs to support their design load requirements will be demonstrated by periodic testing.
Therefore, there is no reduction in the margin of safety associated with this change.
( SORC 92-009)
"Safeguards Equipment Cabinet Control Electronics Unit (CEU) Replacement" Rev. 1 -
The purpose of this DCP is to replace the existing Control Electronics Unit (CEU) in the Safeguards Equipment Cabinet (SEC).
Add a test panel to each of the SECs to facilitate monthly functional tests and 18 month timing tests.
Add three undervoltage test switches to each of the SEC cabinets so as to eliminate the need for jumpering during surveillance testing of the 4KV vital busses.
Add a Diesel Generator start pushbutton to the existing control panel in the SEC cabinets to facilitate testing.
The reliability of the SEC system is demonstrated by analysis/test results that ensure the modified equipment will withstand the affects of a seismic event.
Reliability is further maintained through the use of MIL-Specification material in the construction of the replacement equipment.
The replacement CEUs interface with existing input and output relays.
No new interfaces are introduced, therefore, the existing redundancy and diversity are maintained.
Therefore, this DCP does not reduce the margin of safety as defined in the bases of the Technical Specifications.
( SORC 92-009)
"Pressurizer Insulation Modifications" - This DCP installs NUKON blanket type fiberglass insulation patches with stainless steel jacket for areas on the Pressurizer and loop seal
10CFR50.99 EVALUATIONS MONTH: -
JANUARY 1992 (Cont'd)
ITEM DCP# 2EC-3125 Pkg. 1 B.
Procedures and Revisions S2.0P-AB.ROD-0004(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 FEBRUARY 10, 1992 J. FEST (609)339-2904 enclosure where existing Metal Reflective Insulation (MRI) insulation is missing.
Patches will be installed with seismic attachments to existing MRI insulation.
UFSAR Section 6.3.2.2 discusses the possibility of insulation material being dislodged and blocking the containment sump screens.
An Emergency Core Cooling System Safety Analysis for Salem per USNRC Regulatory Guide 1.82, Revision 1 was performed considering the NUKON insulation being installed by this DCP.
The analysis concludes that the NUKON insulation system will meet the requirements of the Reg. Guide when installed in Salem Unit 2 containment.
Because the proposed patching with he NUKON insulation is to cover gaps in the existing insulation with equivalent functional value, there is no change to the basis of the Technical Specifications.
(SORC 92-010)
"Lube Oil Flushing Modifications" - This change involves the installation of a flushing connection to the main turbine oil reservoir.
The flushing connection will only be used with the plant in the shutdown mode.
Since this change is for the installation of a connection made to the reservoir to facilitate flushing which will be accomplished with the plant in the shutdown mode, it will have no affect on accidents or malfunctions during plant operation.
It also will have no affect on accidents or malfunctions analyzed for the shutdown modes.
Therefore, there is no reduction in the margin of safety as defined in the bases foi any Technical Specifications.
(SORC 92-010)
"Rod Position Indication Failure" Rev. 0 -
The purpose of this procedure is to provide the direction necessary to comply with Technical Specification boration requirements for failed rod position indicators.
This revision changed the old procedure number AOP-ROD-5, Rev. 0 to Sl.OP-AB.ROD-0004(Q), Rev. 0 as part of the Procedure Upgrade Project.
The procedure title
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1
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10CFR50.?9 EVALUATIONS MON~H: -
JANUARY 1992 (Cont'd)
ITEM S2.0P-AB.RHR-0002(Q)
TS2.0P-SO.ZZ-0009(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 FEBRUARY 10, 1992 J. FEST (609)339-2904 was changed to encompass additional guidance for failure of Group Step Counter indication.
This procedure does not create the possibility of an accident or malfunction of a different type than any previously evaluated in the SAR, nor does it increase the probability or consequences of an accident previously in the SAR.
No changes to the the control rod drives has been made.
The method of monitoring once a suspected misaligned control rod is identified is being replaced with the Technical Specification requirement that results in a more detailed method of determining rod position.
( SORC 92-001)
Loss of RHR at Reduced Inventory" Rev. 0 -
The purpose of this procedure is to provide guidance for an.RHR System Malfunction with Reactor Vessel water level below the 101' elevation.
No sections of the UFSAR contain an analysis of Mid-Loop Operation or loss of RHR while at Mid-Loop conditions.
Safety Evaluation S-C-R200-MSE-0738-1 establishes the basis for Mid-Loop Operation.
Responses to Generic Letter 87-12, Loss of Decay Heat Removal, Generic Letter 88-17, Loss of Decay Heat Removal, WOG 90-067, Loss of RHR while operating at Mid-Loop Condition, PSE-90-0557, Salem Mid-Loop analysis and WOG ARG-li and Westinghouse Owners Group Abnormal Response Guidelines provides the bases for the actions in this procedure. (SORC 92-003)
"Removing/Returning 2B 125VDC Bus From/To Service" Rev. 0 -
This procedure provides instruction for removing 2B 125VDC Bus from service and subsequently restoring it to service again.
Redistribution of 2A-125VDC and 2B VDC loads will: 1) maintain adequate DC power to connected loads; 2) not increase battery sizing requirements; 3) not cause a LOPA previously evaluated in the SAR; 4) still allow the batteries to operate as designed, and 5) still allow all safety related systems that use this 125VDC power to operate.
The margin of safety as defined in the bases for the Technical Specification is not reduced.
(SORC 92-005)
I I
1- -- - -
I 10CFR50.99 EVALUATIONS MONTH: -
JANUARY 1992 (Cont'd)
ITEM S2.0P-AB.SW-0001(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 FEBRUARY 10, 1992 J. FEST (609)339-2904 "Loss of Service Water Header Pressure" Rev. 0 -
This procedure is designed to provide the direction necessary for determining and correcting the cause of an abnormal reduction in 21 and/or 22 Service Water header pressure.
There is no reduction in the margin of safety as defined in the bases for any Technical Specification.
The guidance in this procedure allows the operator to use his systems knowledge to determine how to isolate a leak.
Allowing flow to continue to both Component Cooling Water (CCW) Heat Exchangers has no impact on equipment important to safety because a leak in the Diesel Generator service water piping can be isolated with or without service water flow to the CCW Heat Exchangers.
( SORC 92-007)
C.
Temporary Modifications (TMOD)
TMR 91-091 TMR 91-090 "Fuel Handling Building Exhaust" -
The purpose of this modification is to supply temporary power to the #22 Fuel Handling Building Exhaust Fan, No. 2 Battery Room Exhaust Fans, and Radiation Monitoring Sampling Enclosure for the Plant Vent Noble Gas and Post Accident Radiation Monitor R45.
The temporary jumpers are limited in use to Modes 5 and 6.
All of the temporary jumpers must be removed and the equipment returned to normal configuration before entering Mode 4.
This TMOD has no impact on the existing Accident Analysis and it also will not create an accident of a different type than previously evaluated in the SAR.
(SORC 92-001)
"Fuel Handling Building Exhaust" -
The purpose of this modification is to supply temporary power to the #21 Fuel Handling Building Exhaust Fan and 2RP4 Status Panel.
The temporary jumpers are limited in use to Modes 5 and 6.
All of the temporary jumpers must be removed and the equipment returned to normal configuration before entering Mode 4.
This TMOD has no impact on the existing Accident Analysis and it also will not create an accident of a different type than previously evaluated in the SAR.
(SORC 92-010)
SALEM UNIT NO. 2
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SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 2 JANUARY 1992 The Unit was out of service for the entire period for the Sixth Refueling Outage.
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REFUELIN~ INFORMATION MONTH: -
JANUARY 1992 MONTH JANUARY 1992 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
- 1.
Refueling information has changed from last month:
YES X
NO 50-311 SALEM 2 FEBRUARY 10, 1992 J. FEST (609)339-2904
- 2.
Scheduled date for next refueling:
NOVEMBER 11, 1991
- 3.
Scheduled date for restart following refueling:
APRIL 15, 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 x
NO 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 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.