ML18096A636
| ML18096A636 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/1992 |
| From: | Shedlock M, Vondra C Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9204210183 | |
| Download: ML18096A636 (20) | |
Text
.-'0 PS~G e Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station April 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 March 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 lje~y yours, Gg~~
Salem Operations cc:
Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA Enclosures 8-1-7.R4 The Energy People 9204210183 920331 PDR ADOCK 05000311 R
PDR 19046 10~~
95-2189 (10M) 12-89
~RAGE DAILY UNIT POWER LE~
Docket No.:
50-311 Unit Name:
Salem #2 Date:
04/10/92 Completed by:
Mark Shedlock Telephone:
339-2122 Month March 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
- . OPERATING DATA REPORT
- Docket No:
50-311 Date:
04/10/92 Completed by:
Mark Shedlock Telephone:
339-2122 Operating Status
- 1.
Unit Name Salem No. 2 Notes
- 2.
Reporting Period March 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)
- 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 744 0
0 0
0 0
0
-2838 0
0 0
0 0
Year to Date Cumulative 2184 91753 0
58616.1 0
0 0
56898.8 0
0 0
130111721. 8 0
59727048
-6739 56861546 0
62.0 0
62.0 0
56.0 0
55.6 100 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 14. 1992 8-1-7.R2
NO.
DATE 0001 03/01/92 1
2 F:
Forced S:
Scheduled DURATION TYPE 1
(HOURS)
REASON2 s
744 Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refueling D-Requlatory Restriction c
UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH MARCH 1992 METHOD OF SHUTTING DOWN REACTOR 4
3 LICENSE EVENT REPORT #
Method:
1-Manual 2-Manual Scram SYSTEM CODE 4
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 04/10/92" Mark Shedlock*
339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE FUEL XX NUCLEAR NORMAL REFUELING 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File CNUREG-0161) 5 Exhibit 1 - Same Source
SAFETY RELATED MAINTENANCE DOCKET NO:
50-311 MONTH~ ~ MARCH 1992 UNIT NAME:
SALEM 2 DATE:
COMPLETED BY:
TELEPHONE:
APRIL 10, 1992 J. FEST (609)339-2904
~---------------------------------------
WO NO UNIT 870815135 2
900509092 2
900716087 2
900810146 2
901207085 2
910702073 2
910829118 2
911203089 2
911206123 2
920116065 2
EQUIPMENT IDENTIFICATION VALVE 24SW2 FAILURE DESCRIPTION:
REMOVE & REPLACE WITH UPGRADED VALVE VALVE 22SJ138 FAILURE DESCRIPTION:
VALVE 22SJ138 OPERATES HARD -
REPLACE VALVE 2VC5 FAILURE DESCRIPTION:
VALVE 2VC5 FAILED LEAK RATE TEST OPEN & INSPECT 21 BAT PUMP FAILURE DESCRIPTION:
REPLACE POWER FRAME VALVE 2SW184 FAILURE DESCRIPTION:
VALVE 2SW184 BONNET RUSTED -
REPLACE VALVE
- 22 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION:
REPLACE 22 CFCU MOTOR VALVES 22MS169 & 171 FAILURE DESCRIPTION:
INSTALL NEW SIZE 60 ACTUATORS 23 CHARGING PUMP FAILURE DESCRIPTION:
23 CHARGING PUMP MOTOR DOES NOT START -
INVESTIGATE 22 REACTOR COOLANT PUMP FAILURE DESCRIPTION:
22 REACTOR COOLANT PUMP COMPONENT COOLING PIPE CORRODED -
REPLACE VALVE 22SW90 FAILURE DESCRIPTION:
22SW90 UPSTREAM PIPING HAS A THROUGH WALL LEAK -
REPAIR I
I
-~I
SAFETY RELATED MAINTENANCE DOCKET NO:
50-311 SALEM 2 MONTH~ -
MARCH 1992 (Cont'd)
WO NO UNIT 920130128 2
920130140 2
920207075 2
920213151 2
920214067 2
920220105 2
920313181 2
920314096 2
920315089 2
920315091 2
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
APRIL 10, 1992 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION 21 REACTOR COOLANT LOOP FAILURE DESCRIPTION:
REPLACE RTD 2TA2757 INSULATION FAILURE DESCRIPTION:
DAMAGED MRI INSULATION -
REPAIR OR REPLACE VALVE 2SW311 FAILURE DESCRIPTION:
2SW311 HEADER PRESSURE CONTROL VALVE -
NO CONTROL -
INVESTIGATE VALVE 2CV55 FAILURE DESCRIPTION:
CHARGING FLOW CONTROL VALVE 2CV55 WILL NOT MOVE -
INVESTIGATE 21 CHILL WATER PUMP FAILURE DESCRIPTION:
MECHANICAL SEAL LEAKING -
REPLACE 2A DIESEL GENERATOR FAILURE DESCRIPTION:
2A D/G HEATER INDICATION -
TROUBLESHOOT & REWORK 22 SAFETY INJECTION PUMP FAILURE DESCRIPTION:
REPLACE 22 SAFETY INJECTION PUMP MOTOR VALVE 23SS94 FAILURE DESCRIPTION:
NO OPEN INDICATION IN THE CONTROL ROOM -
REWORK AS REQUIRED VALVE 24BF40 FAILURE DESCRIPTION:
VALVE ONLY OPENS TO 80% -
REWORK AS REQUIRED VALVE 22BF19 FAILURE DESCRIPTION:
VALVE EXCEEDED STROKE TIME -
REWORK AS REQUIRED I
SAFETY RELATED MAINTENANCE DOCKET NO:
50-311 SALEM 2 M,O!;JTH £ """ MARCH 19 9 2
{Cont'd)
WO NO UNIT 920315099 2
920315114 2
920319113 2
920320167 2
920323227 2
920325204 2
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
APRIL 10, 1992 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION VALVE 2SV1475 FAILURE DESCRIPTION:
REPLACE SOLENOID VALVE VALVE 24SJ174 FAILURE DESCRIPTION:
VALVE 22SJ174 HAS TUBING LEAK -
INVESTIGATE & REWORK RADIATION MONITOR 2R18 FAILURE DESCRIPTION:
LIQUID WASTE MONITOR 2R18 FAILED INVESTIGATE VALVE 22DA23C FAILURE DESCRIPTION:
TURBO BOOST VALVE STUCK OPEN -
TROUBLESHOOT VALVE 22SW9 FAILURE DESCRIPTION:
22SW9 UPSTREAM SPOOL 1-MP-4099A THROUGH WALL LEAK -
REPAIR 2ABV8 DAMPER FAILURE DESCRIPTION:
2ABV8 DAMPER DOES NOT OPEN -
TROUBLESHOOT & REWORK
10CFR50.59 EVALUATIONS M.Ol';TTH :- '- MARCH 19 9 2
- DOCKET NO:
UNIT NAME:
Dl>i.TE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 APRIL 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# 2SC-2275 Pkg. 1 DCP# 2EC-3039 Pkg. 1 "Relocating the Existing Shroud Doors from Present Location" Rev. 1 -
The purpose of this change is to relocate the existing shroud doors from their present location.
These doors were originally installed as part of DCP 2SM-0442.
The existing doors apparently were not installed in accordance with the design locations.
The existing shroud door locations do not provide adequate visibility to perform inspections of the seal welded penetrations on top of the RPV head.
These inspections are part of the response to NRC Generic Letter 88-05 concerning boric acid corrosion of carbon steel pressure boundary components.
The proposal involves removing the existing doors, sealing the existing openings, providing new openings in the shroud at locations in accordance with the existing design drawings, and reinstalling the existing doors.
The information provided in DCP 2SM-0442 was used as the basis to develop this change package.
This change shall be performed during the 6th refueling outage.
The shroud will be removed to low dose area for ALARA and house keeping reasons.
The only change from the package which was previously reviewed and approved is the revised location of two (2) doors.
{SORC 92-026)
"22A and 22B Condenser Hotwell Levels" -
The purpose of this design change is to correct the discrepancy between 22A and 22B condenser hotwell levels.
The field modifications involve interchanging the control room indicator leads in the Unit #2 Relay Room and the software nomenclature for 22A and 22B hotwells on the P250 Plant Computer.
The condenser hotwell level measurement instrumentation is not safety related and does not interface with any safety related equipment.
The secondary systems that involve the condensers and hotwell level equipment are not important to safety.
10CF.R50. 59 EVALUATIONS M.O~TTH: ~ MARCH 1992 (Cont'd)
ITEM DCP# 2EC-3110 Pkg. 1 DCP# 2EC-3124 Pkg. 2 e
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 The plant can reject heat without the condensers.
This change does not affect any setpoints or process limits.
{ SORC 92-029)
"Allowable Value and Setpoint for Containment High-High (High 2) Pressure" -
The purpose of this change is to establish the new Allowable Value and Setpoint for Containment High-High
{High 2) Pressure based on the reduced Safety Analysis Value of 17 PSIG.
This will result in a change to the Technical Specification Allowable Value from 24.0 PSIG to 16.0 PSIG and therefore a new Trip Setpoint from 23.5 PSIG to 15.0 PSIG based on the setpoint calculation.
This modification will not reduce the margin of safety as defined in the basis for any Technical Specification.
The operation of the syst~ms initiated by this setpoint is fundamentally unchanged from currently described in the UFSAR and safety margins are not reduced.
Setpoint Calculation SC-CS002-03 demonstrates the margin of safety is not reduced since the Allowable Value {AL) -
Nominal Trip Setpoint {NTSP) >
Total Loop Allowance (TLA}.
( SORC 92-029)
"Component Modifications to be Compatible with Replacement Generator" -
The purpose of this change is to address various components which require modifications to be compatible with the replacement generator.
Changes to be accomplished include: A) Removal and replacement recorders 2TA6677R, 2TA6678R and 2TA6679A in Panel 969-2 with three new recorders having the same component identifications but with a new model numbers; B) Relocation of the raceway for the cold gas temperature (GTG-1) from the east side of the old generator to the west side of the new generator, and C} Replacement of hydrogen cooler discharge header temperature controller No. 2TA1782.
The old Salem-2 generator hydrogen coolers No. 1 and No. 2 have a hot gas and a cold gas RTD for a total of four RTDs:
73, 74, 75 and 76.
The new Salem-2 generator is supplied with two RTDs for each hydrogen cooler No. 1 and No. 2 hot gas and cold gas monitors for a total of eight RTDs: 73, 74, 75, 76, 77, 78, 79 and 80.
10CFR50.59 EVALUATIONS M.Ol~TH: '- MARCH 19 9 2 (Cont'd)
ITEM DCP# 2EC-3119 Pkg. 1 DCP# 2EC-3120 Pkg. 1 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 The design documents are revised to ~eflect the four active RT: 73, 75, 77, and 79; the remaining RTDs, 74, 76, 78, and 80 are inactive.
Required splice boxes were not installed on the new generator, resulting in the need for design and installation of two new boxes.
Cable 2TSIS11-FQ was damaged in the area under the generator.
Since a replacement cable is not available, a splice box is being added on El. 120' -
O".
G)
A Thermocouple Conduit Extension to the high-voltage bushing is being installed.
A review of the modifications indicates that all are minor in terms of the existing design.
In addition, the main generator and supporting systems addressed by this DCP are non-safety related.
As such, it is concluded that implementation of the stated modifications will have no affect on the safe operation or shutdown of the plant.
(SORC 92-029)
"Addition of a Parallel Pair of 6 Micron Filters" -
The purpose of this change is to add a parallel pair of 6 micron filters, associated valving, pressure indicators, differential pressure indicating switches, and control room annunciation in the Autostop oil system to prevent debris in the lube oil from clogging the AST orifice.
No change is made to the design or functions of the turbine trip system.
although valves, included for the valving in and out of service of the filter assemblies create a new possibility for the turbine trip, administrative controls, control room alarm, gages, and procedures are put in place to insure that the probability of this type of failure has not been increased.
No safety related functions or systems are impacted.
(SORC 92-030)
"New ASO Gauge to be added to the Turbine Front Standard Panel" - This modification will add a new ASO gauge to the Turbine Front Standard Panel to inform the operator of test header pressure decay and that the header is repressurized after reset at the conclusion of the test.
The pressure gauge rack is located inside the high pressure housing.
10CFR50. 59 EV.i\\.LUATIONS MONTH~ -
MARCH 1992 (Cont'd)
ITEM DCP# 2EC-3125 Pkg. 2 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 The existing gauge order doesn't match the positions of the test valves.
A rearrangement and enlargement of the rack is needed to enhance operator interpretation and control of the AST (Auto Stop System) teat gauges and valves respectively.
The existing orientation of pressure gauges located on the Turbine Standard Front Panel will be rearranged by this modification to match the existing test valve positions.
The rack inside the housing of the Turbine Standard Front Panel is being extended upward in support of two Autostop Oil Pressure Gauges (1 new and 1 existing).
Highly readable nameplates shall be attached to the body of the rack.
This modification will reduce the probability of a turbine/reactor trip by reducing the possibility of closing the test valve before the header is fully repressurized.
This modification will also reduce the probability of an inadvertent trip caused by misinterpretation of which gage is being read.
None of these gages affect the system oil pressure so signals around the setpoint will not be affected.
(SORC 92-030)
"LP Turbine Rotor Replacement" - This DCP involves the documentation of the replacement of the existing low pressure turbine rotor(s) damaged in the turbine overspeed event on November 9, 1991.
The damaged rotors included the heavy disc/keyplate design for discs 1 through 4 on all 3 low pressure elements.
The replacement rotors for LP22 & LP23 include the heavy disc/keyplate design for discs 1 through 3 only.
The replacement LP21 rotor is the original light disc rotor from Salem Unit No.
- 2.
This modification does not reduce the margin of safety as defined in the basis for any Technical Specification.
The Westinghouse methodology in the PRA is not affected.
This modification has no impact on any instrument availability or operability requirements.
all associated control systems and setpoints remain the same and are not impacted by this modification.
(SORC 92-031)
10C1R50.59 EVALUATIONS MONTH: *- MARCH 1992 (Cont'd)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 ITEM
SUMMARY
DCP# 2EC-3126 Pkg. 1 DCP# 2EC-3121 Pkg. 1 DCP# 2EC-3124 Pkg. 12 "Exciter Housing Fire Detectors" -
The exciter housing fire detectors 2FPSD84-1 & 2FPSD84-2 were changed by Temporary Modification (TMOD)89-078 from rate of rise/fixed temperature type to rate compensated/fixed temperature type.
This DCP incorporates the previously installed TMOD into the design documents.
It also replaces the rate compensated fixed temperature detectors subjected to the 11/09/91 turbine generator event with new Fenwal detectors of the same type and model.
The fire detection system modified by this DCP is not part of any Technical Specification, therefore this DCP does not reduce the margin of safety as defined in the basis for any Technical Specification.
(SORC 92-032)
"Installation of Pressure Switches, Solenoid Valves, and.Selector Switches" -
The purpose of this change is to install pressure switches and relays to provide 2/3 logic for 20/ET actuation, install 20-2/AST solenoid valve to ensure additional AST 'redundance during hydraulic testing, and install keyed selector switches and relays to facilitate independent testing of each AST and OPC solenoid valve.
This modification improves turbine trip protection performance.
Administrative controls, supervisory lights, key lock switches and procedures ensure that the probability of an accident has not been increased.
These modifications do not alter the design intent of the turbine trip protection system but increases its reliability.
A malfunction of equipment modified by this DCP could only result in inadvertent turbine trips which is evaluated in chapter 15 of the UFSAR.
(SORC 92-034)
"Separate Cooling Water Supply and Return Lines for the Neutral Tie" Rev. 1 -
The purpose of this change is to provide separate cooling water supply and return lines for the Neutral Tie located below the Generator Bushing Box.
It also makes provision for manual adjustment of cooling water supply pressure to the Stator Bushing Box and Neutral Tie supply lines.
10C'FR50.59 EVALUATIONS MO~ITH; *- MARCH 1992 (Cont'd)
ITEM DCP# 2EC-3078 Pkg. 1 e
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 Low flow alarms from 2PD7359 (63-P98) currently exist for low Rectifier and Bushing/Neutral Tie flows respectively.
Revision 1 to this DCP adds two (2) additional inputs to the turbine runback circuit for low low cooling water flow to the stator bushing and rectifier.
The existing Generator Stator Water Turbine Runback scheme is based on the fact that under certain circumstances it is mandatory that load be removed from the generator in order to protect it from damage.
In case of the stator winding cooling system, a condition of a cooling deficiency would require immediate reduction of armature current to prevent damage to the winding.
Load can be removed by either a trip of the generator breakers, or manipulation of the turbine controls to runback the T-G set.
Since the stator winding does have limited capability with no water cooling, a load runback to the capability limit is often preferred over complete removal of load.
( SORC 92-034)
"Replacement of 27 Rosemount Electronic Transmitters" -
The purpose of this change is to replace 27 Rosemount transmitters with new transmitters having improved performance under harsh environmental conditions.
The transmitters are: 4 Pressurizer pressure; 3 Pressurizer level;12 Steam Generator narrow range level, and 8 Steam Generator steam flow.
The new transmitters are seismically and environmentally qualified for the installed locations.
The tubing and valving arrangements will be standardized for all transmitters of the same type.
Each transmitter will be equipped with a qualified electrical connector to facilitate maintenance.
Where space is available, differential pressure transmitters will be equipped with calibration volume chambers allow calibration with pneumatic signals without the incursion of air into the transmitters or impulse tubing.
There is no change to the instrument process variable spans.
The new transmitters are equal to or better than the previous instruments under all conditions.
10CFR50.59 EVALUATIONS MONTH: *- MARCH 1992 (Cont'd)
ITEM DCP# 2EC-3100 Pkg. 1 DCP# 2EC-3136 Pkg. 1 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 Verification of the P-11 allowable value, consistent with the setpoint calculation, will continue to ensure that manual block of safety injection is defeated well above the SI actuation allowable value of ~ 1755 psi.
There is no impact on the margin of safety as defined in the basis for any Technical Specification.
(SORC 92-031)
"Replacement of RTDs" -
The purpose of this change is to replace the existing RTDs with the new Weed Model No. N9004E-2A-SP and N9259E-2A-SP RTDs.
The new RTDs are the same model as the existing RTDs however a new manufacturing process provides improved insulation resistance and there will b~ a butt splice in the RTD head in place of a terminal post.
The new RTDs are supplied with an EGS model, bayonet style, quick disconnect in place of the Conax quick disconnect.
This will facilitate quicker installation and maintenance of the RTDs thereby reducing exposure time spent in radiation areas.
The performance and accuracy of the new RTDs is identical to the existing RTDs.
time response testing will be performed prior to declaring the new system operable to verify that the new RTDs exhibit acceptable results.
Based on the above, the proposed modification does not reduce the margin of safety as defined in the basis of any Technical Specification.
(SORC 92-031)
"Replacement of Failed Lambda Power Supplies and Daisy Chain" -
The purpose of this change is to replace two failed Lambda power supplies with equivalent Lambda power supplies for all five Rod Control Cabinets in Unit 2 and replace existing daisy chain neutral wiring between the power supplies in a Rod Control Cabinet with individual neutral wires for each power supply.
The rod control circuits which are powered by the new power supplies are not safety related.
The replacement power supplies are equal to or exceed the existing power supplies in all physical and electrical parameters.
10CFR50.59 EVALUATIONS MONTH: *- MARCH 19 9 2 (Cont'd)
ITEM DCP# 2EA-1022 Pkg. 1 DCP# 2SC-2232 Pkg. 1 e
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 The separate neutral wiring from each of these power supplies to the neutral bus will minimize inadvertent tripping of the reactor during changes to any individual power supply when the reactor is at full power.
(SORC 92-036)
"Reclassification of Piping Specification Piping Schedule SPS53D from Nuclear Class III to Non-Nuclear Class" -
The purpose of this change is to reclassify piping specification Piping Schedule SPS53D from Nuclear Class III to Non-Nuclear.
Changes are being made to numerous P&IDs piping class classification tables to incorporate the above changes.
Additional piping class break clarifications and corrections are being made as well.
The other document updates required to incorporate the above changes are UFSAR, S-C-MPOO-MGS 0001-SPS53
& MMIS.
Engineering Evaluation S-C-WD-MEE-0692 documents and accepts the above Nuclear classification changes.
The margin of safety for the Technical Specifications is not reduced since the piping class 53D that is being reclassified is already non-safety related and not part of the Technical Specification bases.
(SORC 92-036)
"Replacement of Two flanged Spool Pieces with Containment Spray System Valves 21CS6 and 22CS6" Rev. 1 -
The purpose of this revision to the DCP is to remove valves 21CS6 and 22CS6 and reinstall the spool pieces.
The valves installed per the initial issue of this DCP are designed for a differential pressure of 225 psi and did not hold pressure at 47 psi.
The CS6 valves are locked open during normal operation.
Both are manual valves.
They are closed for Containment Spray (CS) pump full flow testing to direct the discharge to the refueling cavity or for leak rate testing to the CS48 valves.
Both of these tests, which are conducted during refueling outages do not affect the normal operation of the Unit or the availability of the CS system as per the Technical Specification requirements.
( SORC 92-037)
10C~R50.59 EVALUATIONS MONTH: *- MARCH 1992 (Cont'd)
ITEM DCP# 2EC-3114 Pkg. 1 DCP# 2EC-3123 Pkg. 1 B.
Procedures and Revisions S2.0P-PT.CS-0103(Q)
DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 "Miscellaneous Turbine Control Systems" - This change: 1) Added a 110% overspeed trip signal to the regular turbine trip circuit with interlock from their Main Generator Breakers; 2) Modified the shape of the turbine overspeed trip test lever; 3) Interlocked the operation of the main steam stop valves.
This modification only adds an additional means of tripping the turbine on an overspeed condition and has no effect on the OPC or mechanical overspeed trip.
( SORC 92-038)
"Reverse Power Relaying Protection" - This modification installs primary and backup reverse power relays interlocked with turbine trip contact K635 for generator protection against motoring above 6 MW of reverse power flow. A 30 second time delay between turbine trip and generator trip is required so that full reactor coolant flow can be maintained to ensure adequate removal of core heat and to prevent turbine overspeed. This change provides generator protection against extended motoring and turbine protection against potential overheating of the low pressure turbine blades when generator motoring occurs.
The proposed modification does not reduce the margin of safety as defined in the Technical Specification because this modification does not impact the Technical Specification or its basis.
No changes to the Technical Specification setpoints or equipment ratings have been made.
(SORC 92-038)
"Differential Pressure Test (No Flow) of Containment Spray Pump Discharge Stop Valves 21CS2 and 22CS2" Rev. 0 -
The purpose of this procedure is to obtain performance data for 21(22)CS2.
This data will be used to determine if the CS2 valves can open under design delta P as required by Generic Letter 89-10.
This test will be performed in modes 5 or 6 with the Reactor Vessel Head installed.
After VOTES has collected acceptable data the pump is stopped, CS2 is shut and the system returned to normal.
I I
I I
10CPR50.59 EVALUATIONS MONTH: *- MARCH 1992
{Cont'd)
ITEM C.
Safety Evaluations {S/E)
NFU 92-181 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 Technical Specifications (TS) Section 3.6.2.1 does not require Containment Spray (CS) to be operable in Modes 5 and 6.
If a Service Water Header is out of service, the CS pumps may be the two required safety grade pumps per TS 3.4.1.4.
If the CS pumps are the designated safety grade pumps, TS 3.4.1.4 will be complied with by having a Safety Injection pump and CS pump not being tested operable.
Therefore, the margin of safety as defined in the TS is not reduced.
( SORC 92-02 8)
"Revision to Salem Unit 2 Cycle 7 Reload Safety Evaluation" -
During the Salem Unit 2 Cycle 6/7 Refueling, 70 fuel assemblies will be replaced with 8 Region 4 fuel assemblies, 9 Region 5A fuel assemblies, 1 Region 6 fuel assembly, 16 fresh Region 9A fuel assemblies (3.8 w/o containing 64 Integral Fuel Burnable Absorbers (IFBA) pins), 8 fresh Region 9B fuel assemblies (4.0 w/o containing 64 IFBA pins), 16 fresh Region 12 fuel assemblies (4.0 w/o containing 64 IFBA pins), and 12 Region 12 fresh Fuel assemblies (4.0 w/o without IFBA pins).
The Cycle 7 core contains 544 fresh burnable absorber rodlets arranged in clusters as shown in Attachment lA.
This evaluation covers all operational modes.
Deficiencies affecting the operation of Cycle 7 were identified in DEF's DES-91-00778 and DES-92-00354 (which involved and increase in Auxiliary Feedwater (AFW) maximum flow rates and an increase in Containment Spray delay time).
As per Safety Evaluation S-C-CS-MSE-0804, NFU-92-095, &
NFU-92-116, the effects that these deficiencies have upon safety analysis inputs were quantified and provided to Westinghouse for incorporation into the Cycle 7 safety analyses.
The results of these evaluations conclude that Salem Unit 2 Cycle 7 reload design will not result in the previously acceptable safety limits for any accident being exceeded and does not result in any unreviewed safety questions.
10C~R50.59 EVALUATIONS MONTH*: **- MARCH 1992 (Cont'd)
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 APRIL 10, 1992 J. FEST (609)339-2904 It has been verified that the postulated containment pressure and temperature do not exceed design limits.
In addition, a reduction in the Containment Pressure HIGH-HIGH setpoint (DCP# 2EC-3110) in conjunction with available excess cycle 7 shutdown margin was credited in this safety evaluation.
( SORC 92-033)
"Operable Service Water Loop" - Technical Specification Interpretation LC0-3.7.4 specifies the Service Water pump operability requirements in order to meet Tech Spec 3.7.4. Technical Specific~tion 3.7.4 indicated that two independent Service Water loops must be operable.
The basis for this Tech Spec states that the operability of the Service Water system ensures that sufficient cooling capacity is available during normal and accident conditions.
The basis also specifies that the redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analysis.
Redundancy in the Service Water system is provided by six pumps and two headers.
To assume sufficient cooling capacity is available during a DBA LOCA, two Service Water pumps are required (reference UFSAR section 9.2.1.2).
(SORC 92-029)
1
SALEM UNIT NO. 2 SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 2 MARCH 1992 The Unit was out of service for the entire period for the Sixth Refueling Outage.
- - REFUELING INFORMATION MONTH:*- MARCH 1992 MONTH MARCH 1992 DOCKET NO:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
- 1.
Refueling information has changed from last month:
YES NO X
50-311 SALEM 2 APRIL 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 193
- b.
In Spent Fuel Storage 408
- 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.
- -I
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