ML18100B127
| ML18100B127 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/31/1994 |
| From: | Hagan J, Heller R, Morroni M Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9406200032 | |
| Download: ML18100B127 (12) | |
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PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 June 14, 1994 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 May 1994 are being sent to you.
RH:pc Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions 10CFR50.59 Evaluations Operating Summary Refueling Information 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 Peopie 9406200032 940531 -- -- -,\\
PDR ADOCK 05000311 R
PDR 95-2189 (10M) 12-89
/,
~ERAGE DAILY UNIT POWER LE~
Docket No.:
50-311 Unit Name:
Salem #2 Date:
06-10-94 Completed by:
Mike Morroni Telephone:
339-2122 Month May 1994 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET)
(MWe-NET) 1 329 17 1052 2
765 18 1036 3
885 19 1056 4
1012 20 549 5
1098 21 463 6
999 22 614 7
1046 23 1031 8
1071 24 1048 9
1066 25 1030 10 1058 26 1008 11 1043 27 908 12 1131 28 1032 13 985 29 1043 14 1114 30 1067 15 981 31 1053 16 1056 P. 8.1-7 Rl
/
OPERATING DATA REPORT ~
Docket No:
Date:
Completed by:
Mike Morroni Telephone:
- operating status
- 1.
Unit Name Salem No. 2 Notes
- 2.
Reporting Period May:
1994
- 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 N A.
This Month 744 744 0
744 0
2335672.8 743330 711102 100 100 86.4 85.7 0
Year to Date 3623 3546.0 0
3462.3 0
10167544.5 3254460 3098446 95.6 95.6 77.3 76.7 4.4 50-311 06/10/94 339-2122 since Last NIA cumulative 110736 72825.6 0
70351. 8 0
172630483.9 73800978 70225723 63.5 63.5 57.3 56.9
- 21. 9
- 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
Refueling outage scheduled to start October 15, 1994 and last 60 da s.
- 25. If shutdown at end of Report Period, Estimated Date of Startup:
NA 8-l-7.R2
NO.
DATE 0838 5-20-94 F
1 2
F:
Forced S:
Scheduled DURATION TYPE 1
(HOURS)
REASON2 18.3 B
Reason A-Equipment Failure (explain)
B-Maintenance or Test C-Refueling D-Requlatory Restriction UNIT SHUTDOWN AND POWER REDUCTIONS 5
REPORT MONTH MAY 1994 METHOD OF SHUTTING DOWN REACTOR LICENSE EVENT REPORT #
CH 3
Method:
1-Manual 2-Manual Scram SYSTEM CODE 4
E-Operator Training & License Examination 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther F-Administrative G-Operational Error (Explain)
H-Other (Explain)
COMPONENT CODE5 DOCKET NO.
UNIT NAME DATE COMPLETED BY TELEPHONE 50-311 Salem #2 06-10-94 Mike Morroni 339-2122 CAUSE AND CORRECTIVE ACTION.
TO PREVENT RECURRENCE PUMP XX FEEDWATER PUMPC ONTROLS 4
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same Source
~.
10CFR50.59 EVALUATIONS MONTH: -
MAY 1994 DOCKET
- UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 SALEM 2 JUNE 10, 1994 R. HELLER (609)339-5162 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 2EC-3274 Pkgs 1&2 "Throttling Service Water Flow To No. 21 and 22 CCHX" - These change packages throttle the Service Water inlet isolation valves from the existing full open locked positions to partially open locked positions.
These particular valve positions will be established in the field and dependent on the existing CCHX's flow resistances.
These packages also address the throttling of Service Water inlet and outlet valves.
The inlet and/or outlet isolation valves may then have to be adjusted to compensate for +/- changes in CCHX flow resistance.
The Component Cooling HX operating procedures will also be revised to update the CCHX design outlet temperatures from 95 degrees to 99 degrees F.
- Also, the compensatory actions imposed by Engineering Evaluation S-C-SW-MEE-0893, for SWS operation with river temperatures < 70 degrees F, will be removed from these procedures.
The Westinghouse analysis demonstrates that the proposed throttling of valve 21SW21, and valve 21SW355, if required, will assure adequate service water flow rate delivery to the CCHX under postulated accident conditions.
Calculation S-C-MDC-1317 shows that the proposed valve throttling during all modes of the plant operation will still ensure the original service water design flow rate of 10,000 gpm through the CCHX during all normal operating conditions.
Thus the acceptance criteria for licensing basis accidents and transients will continue to be satisfied and the existing safety margins will not be affected.
(SORC 94-043) 2EC-3276 Pkg 1 "Relay Coordination for the 4160/480-240V Vital Transformers" - This DCP changes the overcurrent relay settings at the 4.16KV vital switchgear compartments 2A4D, 2B4D, 2C4D, 23 & 24ASD, 23 &
24BSD and 23 & 24CSD.
The implementation of this DCP will not involve changes of any kind to plant
10CFR50.59 EVALUATIONS MbNTH: -
MAY 1994 (cont'd)
ITEM 2EC-3265 Pkg 1 2EX-2137 Pkg 1
SUMMARY
DOCKET N' UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-272 SALEM 1 JUNE 10, 1994 R. HELLER (609)339-5162 processes and it will not create the possibility of an accident of a different type than previously evaluated in the SAR.
This DCP will not increase the consequences of a malfunction of a different type than any previously evaluated in the SAR.
The relay setting changes to the overcurrent relays of the 4.16KV vital bus incomer feeder and 4160/480-240 transformers will not reduce the margin of safety as defined in the basis for any Technical Specification.
(SORC 94-045)
"Deletion of RMS Channels 2R35 and 2R38" -
The purpose of this DCP is to perform modifications to remove Channels 2R35 and 2R38 from the RMS.
Detectors, check sources, and monitors are to be removed from the field.
Cables, conduits, and conduit supports are to be removed, rerouted, or spared in place.
2R35 and 2R38 are to be physically removed from the Interdata 7/32 communication loop.
The monitors to be removed are in the Steam Generator Blowdown Filter Loop, which is blind flanged out of service.
RMS Channels 2R35 and 2R38 are not included in the Technical Specifications, nor in the Bases for the Technical Specifications.
The monitors perform no function to mitigate radiation exposures at the site boundary, resulting from accidents described in the UFSAR, chapter 15.
Therefore, this modification does not reduce the margin of safety as defined in the bases for any Technical Specification.
(SORC 94-045)
"Caldon Ultrasonic Flowmeter Installation Test &
Experiment" -
The proposed modification to the facility is the installation of flow measurement test equipment on each of the four f eedwater pipes to the steam generators, and data recording equipment on the P-250 plant computer.
This equipment is being installed for the purpose of determining the current f eedwater flow rate to the steam generators, determining the present conditions of the feedwater flow nozzles, and providing a mechanism to trend performance and degradation of the system.
The reason for performing this test is
10CFR50.59 EVALUATIONS MONTH: -
MAY 1994 (cont'd)
.ITEM
SUMMARY
DOCKET :.I UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
50-272 SALEM 1 JUNE 10, 1994 R. HELLER (609)339-5162
~--------------------------------------------------------------------
that f eedwater flow measurement was determined to be questionable as a result of review to Unit 2 Fuel Cycle 8 calorimetric and reactor Coolant System flow calculations.
This equipment will not be used as a primary method of controlling feedwater flow at this time.
Failure of the equipment will only result in the loss of data for the period of time the equipment is failed.
The Unit 2 Technical Specifications are primarily concerned with steam flow and feedwater flow mismatch with respect to steam generator water level control.
The proposed ultrasonic instrumentation is non-intrusive to the f eedwater mechanical and controls system and only provides data retrieval, storage and analysis.
However, the Honeywell transmitters will be installed in an intrusive manner, as they will be connected in parallel to existing transmitters.
The installation of these transmitters is controlled in a manner similar to the existing calorimetric transmitters, and will be implemented such that adverse effects on the f eedwater control and protection systems are minimized.
There are no changes to the facility introduced by the installation of either the ultrasonic instrumentation or the Honeywell transmitters used in this testing modification which would alter the performance of the f eedwater system as described in the Technical Specifications.
As such, the margin of safety has not been reduced. (SORC 94-045)
B. Temporary Modifications T-MOD #94-043 "Unit 2 Fuel Handling Building Ventilation System".
DCP 2EC-3242 provides design changes to the Fuel Handling Building exhaust fans 2VHE20 and 2VHE21, including (1) the replacement of backdraft dampers, flexible connections and adjustable motor sheaves, (2) the installation of duct mounted access doors on fan inlet ducts.
Normally both exhaust fans operate continuously along with the system supply fan 2VHE24. However, during the installation of DCP 2EC-3242, the Fuel Handling Area ventilation system
10CFR50.59 EVALUATIONS MONTH: -
MAY 1994 (cont'd)
ITEM
- c.
Evaluations S-2-RCP-EEE-0899 DOCKET N' UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 JUNE 10, 1994 R. HELLER (609)339-5162 will be taken out of service.
This T-MOD provides blanking plates on the ductwork downstream of the exhaust fan discharge backdraft dampers to ensure gaseous effluent from the Auxiliary Building Ventilation, Containment Pressure Relief and/or Waste Decay Tank does not'inadvertently enter Auxiliary Building, Elev. 100 1, while the required work is being performed as outlined in DCP 2EC-3242.
The Technical Specification requires that whenever irradiated fuel is in the storage pool, the fuel handling area ventilation system shall be operable.
In the event the system is not operable, all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool shall be suspended until the ventilation is restored.
During the temporary modification all fuel handling operations will be suspended in accordance with Technical specification 3/4.9.12 and procedure S2.0P-SO.FHV-00l(Q) and therefore the temporary modification will not requce the margin of safety as defined in the basis of the Technical Specifications.
(SORC 94-042)
The purpose of this proposal is to accept the discovered condition as-is, until the next refueling outage or forced outage of sufficient duration to correct with a design change package.
Each signal summator, 2PM-505B and 2PM-506B, uses the turbine impulse pressure signal to provide the setpoint to one protection channel of High Steam Line Flow Safety Injection (SI) comparators.
Signal summator PM-505B provides the setpoint signal for comparators FC-512, FC-522, FC-532, and FC-542 (Channel I).
Signal summator PM-506B, provides the FC-543 (Channel II, 221056).
These comparators provide the High Steam Line Flow SI signal to Reactor Protection Trains A and B.
The setpoint is reduced to 40%
steam flow once a reactor trip is sensed.
SI initiation results from High steam Line Flow (1/2 taken twice for four loops) together with either Low Low Tave to Low Steam Line Pressure.
If these
10CFR50.59 EVALUATIONS MbNTH: -
MAY 1994 (cont'd}
ITEM D.
Deficiency Report SMD-94-056/057 DOCKET N' UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 JUNE 10, 1994 R. HELLER (609)339-5162 summators lose power, upon re-energizing these modules, the module output saturates high.
This output signal then provides a high steam flow reference setpoint to the High steam Line Flow comparators associated with this particular summator.
With this high setpoint, the associated comparators will never "see" a high steam line flow condition since the setpoint would be significantly greater than 110% steam flow value as defined by Technical Specifications.
The saturated summator output would also be prevented from reverting to the 40% value following a reactor trip.
This, in essence, removes all protective action provided by this summator and thus one of the two High Steam Line Flow protective channels for the period of time require by the summator to come out of saturation.
Both high steam flow channels are verified to be operable by quarterly channel functional testing.
In order to lose both channels of High Steamline Flow protection, multiple failure of equipment required operable by Technical Specifications would have to occur.
Since accident analyses are based on a single failure plus initiating event, the margins of safety demonstrated by the accident analyses remain valid.
(SORC 94-038}
"Plug, Stem and Cage Trim Assemblies" -
During the Unit 2 Seventh Refueling Outage, plug, stem and cage trim assemblies composed of 17-4 PM material were inadvertently installed in the Salem Unit 2 PORVs under DCP 2EC-3190.
The original intent of the DCP was to replace the installed trim assemblies composed of 17-4 PH stem and cage material and 304 SS-Stellite plug material with new trim assemblies composed of 420 SS plug and cage material and chrome-plated 316 SS stem material.
The purpose of this evaluation is to examine the existing condition and evaluate its acceptability for continued operation for the remainder of the current operating cycle.
The bases for Technical Specification 3/4.4.5 states that the PORV's function to relieve pressure during all design transients up to and
10CFR50.59 EVALUATIONS MONTH: -
MAY 1994 (cont'd)
ITEM DOCKET..
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
SUMMARY
50-311 SALEM 2 JUNE 10, 1994 R. HELLER (609)339-5162 including the design step load decrease with steam dump and to minimize undesirable opening of the pressurizer safety valves.
The bases for Technical Specification 3/4.4.10.3 states that the operability of the POPS ensures that the RCS cold legs are less than or equal to 212°F.
The PORVs will perform as designed with reasonable assurance and reliability and will remain capable of performing their specified functions for the current operating fuel cycle.
As a result, we conclude that the existing condition does not reduce the margin of safety as defined in the basis for any Technical Specification.
(SORC 94-041)
SALEM GENERATING STATION MONTHLY OPERATING
SUMMARY
UNIT 2 MAY 1994 SALEM UNIT NO. 2 The Unit began the period operating at reduced power to facilitate replacement of a failed Reactor Coolant System flow transmitter.
Power was increased to 75% on May 2, 1994 following completion of repairs.
Power was further increased to 100% on May 4, 1994 following completion of repairs to a condensate pump seal.
Power was reduced to 50% on May 20, 1994 to repair #22 steam generator feed pump turbine speed bias control.
The repairs were completed and power returned to 100% on May 22, 1994.
With the exception of a minor load reduction on May 27, 1994, to clean 23B condenser waterbox, the Unit continued to operate at 100% power throughout the remainder of the period.
REFUELING INFORMATION M'ONTH: -
MAY 1994 MONTH MAY 1994 DOCKET I:
UNIT NAME:
DATE:
COMPLETED BY:
TELEPHONE:
- 1.
Refueling information has changed from last month:
YES NO ~-X=-~-
- 2.
Scheduled date for next refueling:
OCTOBER 15. 1994 50-311 SALEM 2 JUNE 10, 1994 R. HELLER (609)339-5162
- 3.
Scheduled date for restart following refueling: DECEMBER 13, 1994
- 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?:
OCTOBER 94
- 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 492
- 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