ML18106A556
| ML18106A556 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/1998 |
| From: | Bakken A, Knieriem R Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LR-N98-0192, LR-N98-192, NUDOCS 9804210435 | |
| Download: ML18106A556 (10) | |
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- Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236
- '.! Nuclear Business Unit U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk MONTHLY OPERATING REPORT SALEM UNIT NO. 1 DOCKET NO. 50-272 Gentlemen:
April 15, 1998 LR-N98-0192 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original Monthly Operating report for March, 1998, is attached.
RBK/rbk Enclosures c Mr. H. J. Miller Sincerely, A. C. Bakken Ill General Manager -Salem Operations Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046 ------9ao421o435-98o33i--
._, PDR ADOCK 05000272 R PDR The power is in your hands: **-'"" -----1! I 95-2168 REV. 6/94
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- Reporting Period: March 1998
- DOCKET NO.: 50-272 UN IT: Salem 1 DA TE: 4/15/98 COMPLETED BY: R. Knieriem TELEPHONE:
(609) 339-1782 OPERATING DATA REPORT Design Electrical Rating (MWe-Net) 1115 Maximum Dependable Capacity (MWe-Net) 1106 Month Year-to-date Cumulative No. of hours reactor was critical No. of hours generator was on line (service hours) Unit reserve shutdown hours Net Electrical Energy (MWH) UNIT SHUTDOWNS NO. DATE TYPE F=FORCED S=SCHEDULED 1 3/1/98 F -3/31/98 (1) Reason A -Equipment Failure (Explain) B -Maintenance or Test C -Refueling D -Regulatory Restriction DURATION (HOURS) 744 E -Operator Training/License Examination F -Administrative G -Operational Error (Explain) H -Other Summary: REASON (1) F,C 0 0 0 0 0 104380 0 100388 0 0 0 100136543 METHOD OF CORRECTIVE SHUTTING ACTION/COMMENT DOWN THE REACTOR (2) 4 Steam Generator Replacement and Refueling Outage (2) Method 1 -Manual 2 -Manual Trip/Scram 3 -Automatic Trip/Scram 4 -Continuation 5 -Other (Explain)
The unit is in a refueling and a steam generator replacement outage and remained shutdown for the entire period. According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, the unit will remain shutdown pending completion of the following actions: -1
- DOCKET NO.: 50-272 UNIT: Salem 1 DATE: 4/15/98 COMPLETED BY: R. B. Knieriem TELEPHONE:
(609) 339-1782
- Appropriately address long standing equipment reliability and operability issues.
- After the work is completed, conduct a restart readiness review to determine for ourselves the ability of the unit to operate in a safe, event free manner.
- After the restart review, meet with the NRC and communicate the results of that review.
- DOCKET NO.: 50-272 UNIT: Salem 1 DATE: 4/15/98 COMPLETED BY: R. B. Knieriem TELEPHONE:
(609) 339-1782
SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE SALEM UNIT 1 GENERATING STATION MONTH MARCH 1998 The following items completed during March 1998 have been evaluated to determine:
- 1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or 3. If the margin of safety as defined in the basis for any technical specification is reduced. The 1 OCFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. The 1 OCFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
Design Changes Summary of Safety Evaluations 1EC-3206, 115 VAC Vital Bus Uninterruptable Power Supply (UPS) Uprate. This modification uprates all four Salem Unit 1 Vital Bus inverters from 1 O KVA to 12 KV A to support the implementation of the Advanced Digital Feedwater Control system. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3238, Pkg. 1, Salem Unit 1 Steam Generator Feed Pump Lube Oil Cooler. This design change replaced the existing cement lined Service Water piping and components for the Steam Generator Feed pump Lube Oil coolers with stainless steel piping and components which will provide greater system reliability.
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- This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3313, Pkg. 1, Pressurizer Enclosure Component Modification.
This modification provided upgrades for components located within the pressurizer enclosure to improve operability, maintainability, and component service life. The modifications:
(1) replaced the pressurizer spray bypass valves, (2) replaced selected environmentally qualified cables, and (3) rerouted control air tubing for Pressurizer Spray valve positioners and electro-pneumatic transducers.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3353, Pkg. 1, Small Bore Piping Replacement.
In order to improve reliability and to reduce maintenance, this modification replaced piping and valves with upgraded material in the following systems: Main, Reheat, and Turbine Bypass Steam Moisture Separator Reheater Steam and Drain Lines Miscellaneous Drains Bleed Steam and Heater Drains Steam Generator Feed Turbine Drains Miscellaneous Condensate Steam Generator Drains and Slowdown This modification also added new isolation valves in several Main Steam system drain lines to allow isolation for maintenance.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3384, Pkg. 1, Replacement Of Main Generator Hydrogen Gas Purity Meter. This design change upgraded the existing analog Hydrogen Purity Transmitter/
Indicator for the Main Generator with a digital microprocessor based system. ! I . I
- This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3363, Pkg. 1, Revision Of Rod Control System Compensation Parameters.
This design change implemented vendor recommended changes to compensation parameters associated with the Rod Control System to prevent false control rod stepping during steady-state operation with the Rod Control system in automatic.
-This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
Therefore, this design change does not involve an Unreviewed Safety Question.
1EC-3419, Pkg. 1, 480 V Power Supply Relocation.
This modification changed the power sources for several non-safety related circuits to enhance the overall availability of Salem Units 1 and 2. The change reconfigured two station power supplies so that a loss of Station Power Transformer T 4 or T3 would only affect one Salem unit and not result in the trip of both units. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Pkg. 1, Elimination Of Nuclear Instrumentation System (NIS) Flux Deviation Spurious Alarms. This modification installed two time delay printed circuit cards in the NIS Flux Deviation and Miscellaneous Control Drawer (Detector Current Comparator Section) in order to eliminate spurious Quadrant Power Tilt Ratio alarms. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3504, Pkg. 2, Nuclear Instrumentation System Grounding.
This design change modified the grounding of each of the four safety-related NIS racks to conform with Westinghouse design requirements.
This design performs the same function as the previously installed grounding system but with less resistivity to ground, more noise immunity, and the establishment of an equipotential ground plane closer to ground mat potential.
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- This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3525, Pkg. 1, Relief Valve 1CV141 Replacement.
This modification replaced the existing 1 CV141 relief valve with a new Crosby Type JRAK-BS relief valve. The new valve is designed to minimized the effect of fluctuating backpressure on the valve caused by changes in pressure in the Volume Control Tank. Variations in backpressure could affect valve relieving capacity, opening and closing pressure, and valve stability.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EC-3679, Pkg. 1, Safety Injection Discharge Pressure Rerating.
This modification rerates a portion of the Safety Injection system to a new design pressure of 1,910 PSIG to allow for a higher setpoint for relief valves 11 SJ39, 12SJ39, and 1SJ167 (from 1,750 to 1910 PSIG). This was accomplished to prevent valve lifting and subsequent damage to those valves due to transient pressure spikes that occur during surveillance testing. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EE-0068, Pkg. 1, Replacement Of Fish Baskets, 12A Circulating Water Traveling Screens. This modification replaced the existing fish baskets in the traveling screens with a design to improve fish survivability.
The modification also added spray nozzles to improve debris removal and to aid fish handling.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EE-0070, Pkg. 1, Replacement Of Fish Baskets, 13A Circulating Water Traveling Screens. This modification replaced the existing fish baskets in the traveling screens with a design to improve fish survivability.
The modification also added spray nozzles to improve debris removal and to aid fish handling.
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- This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1EE-0102, Pkg. 1, Pressurizer Safety Relief Valve (SRV) Testing. This modification installed strain gauges on the piping associated with the Pressurizer SRVs. The strain gauges were installed to determine the pipe moments acting on the pressurizer SRVs to evaluate the impact of such moments on SRV leakage. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1 EE-0152, Pkg. 1, Centrifugal Charging Pump Flow Control Valve Modification.
This design change replaced the Centrifugal Charging Pump Flow Control valve with a valve that will provide more reliable flow control during normal and depressurized Reactor Coolant system operation.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
1EE-0367, Pkg. 1, Salem Unit 1 Steam Generator Replacement Project: Instrument Rescaling And Recalibration.
This design change addressed the rescaling and recalibration of Salem Unit 1 level instrumentation that was required because of the differences in detector piping nozzle elevations between the old and the replacement Steam Generators.
This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Minor Modification, S97-127, Pkg. 1, 2, 3, and 5, Salem Unit 1 Service Water Flow Element, Instrument Tubing/Root Valve Addition.
This modification added root valves and instrument taps to flow orifice flanges in the Service Water system. The additional monitoring capability will allow for the measurement and monitoring of Service Water system flows to individual safety related components to ensure that the Service Water flows to those components are within the design basis of the plant.
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- This design change does not negatively impact any accident response.
This change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
Therefore, this design change does not involve an Unreviewed Safety Question.
Minor Modification, S97-145, Installation Of An Isolation Valve Downstream Of Valve. This design change installed a non-safety related manual isolation valve to the Condensate system downstream of air operated valve 1 CN56. The new valve will permit isolation of valve 1 CN56 from piping common to Unit 2 to allow for maintenance of valve 1 CN56 while the Unit 2 hotwell level control system is in service. This design change does not negatively impact any accident response.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
Therefore, this design change does not involve an Unreviewed Safety Question.
Temporary Modifications Summary of Safety Evaluations There were no changes in this category implemented during March, 1998. Procedures Summary of Safety Evaluations There were no changes in this category implemented during March, 1998. UFSAR Change Notices Summary of Safety Evaluations UFSAR Change Notice 98-011, Control Room Ventilation (CRV) -Radiologically Contaminated Air Intrusion.
This UFSAR Change Notice documented a change to the UFSAR description of CRV system operation for the Control Room Emergency Air Conditioning system (CREACS) in the single filtration train alignment, without pre-selecting an outside air intake. This evaluation does not negatively impact any accident response.
This procedure revision does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this
- procedure revision does not involve an Unreviewed Safety Question.
Deficiency Reports . Summary of Safety Evaluations There were no changes in this category implemented during February, 1998.
- Other Summary of Safety Evaluations Safety Evaluation S97-026, Offsite Dose Calculation Manual, Revision 11. Revision 11 to the Offsite Dose Calculation Manual (ODCM) corrected Safety Factor values used to calculate radioactive effluent waste release rates and radioactive liquid waste radiation monitor setpoints.
This evaluation does not negatively impact any accident response.
This procedure revision does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore; this procedure revision does not involve an Unreviewed Safety Question.