ML18107A531

Monthly Operating Rept for Aug 1999 for Salem,Unit 1.With 990913

ML18107A531
Person / Time
Site:	Salem
Issue date:	08/31/1999
From:	Bezilla M, Knieriem R Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N99-0415, NUDOCS 9909170017
Download: ML18107A531 (5)

Text

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e PS~G Public Service ElecJric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit

• SEP.131999 LR-N99-0415 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:

In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original Monthly Operating report for August 1999 is attached.

Sincerely,

~~¥ Vice President - Operations

/rbk Enclosures C Mr. H. J. Miller Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046

-99o9i7ooi799oa3i_____ -------- ,,

PDR ADOCK 05000272 )

R PDR )

The power is in your hands.

95-2168 REV. 6/94

DOCKET NO.: 50-272 UNIT: Salem 1 DATE: 9/15/99 COMPLETED BY: R. Knieriem TELEPHONE: (856) 339-1782 Reporting Period: August 1999 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 744 5691 116511 No. of hours generator was on line (service hours) 744 5658 112246 Unit reserve shutdown hours 0 0 0 Net Electrical Energy (MWH) 738401 5972206 112584369 UNIT SHUTDOWNS NO. DATE TYPE DURATION REASON METHOD OF CORRECTIVE F=FORCED (HOURS) (1) SHUTTING ACTION/COMMENT S=SCHEDULED DOWN THE REACTOR (2)

(1) Reason (2) Method A - Equipment Failure (Explain) 1 - Manual 8 - Maintenance or Test 2 - Manual Trip/Scram C - Refueling 3 - Automatic Trip/Scram D - Regulatory Restriction 4 - Continuation E - Operator Training/License Examination 5 - Other (Explain)

F Administrative G- Operational Error (Explain)

H-Other Summary:

Salem Unit 1 began the month of August 1999 operating at full power. Full power operation continued until August 13, when the unit entered end-of-cycle power coast down toward its 131h refueling outage, scheduled to begin on September 17, 1999.

Salem Unit 1 continued to coast down until August 20, when the unit experienced a runback to 47% power following the loss the 11 Steam Generator Feed Pump due to a fault in an Essential Controls Inverter. Salem Unit 1 returned to its maximum attainable power of 88% on August 26. The unit continued to coast down through the remainder of the month, completing the month at a maximum attainable power of 85%.

DOCKET NO.: 50-272 UNIT: Salem 1 DATE: 9715799 COMPLETED BY: R. 8. Knieriem TELEPHONE: (856) 339-1782

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE SALEM UNIT 1 GENERATING STATION MONTH: August 1999 The following items completed during August 1999 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 10CFR50.59 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 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.

Design Changes - Summary of Safety Evaluations Modification 1 EE-0391, Pkg. 1, Addition of Isolation Capability to Spent Fuel Pool Tell-Tale Leak-Off Lines This modification provides isolation capability to the Spent Fuel Pool (SFP) tell-tale leak-off lines to preclude a potential SFP inventory loss in the event that cooling is lost and the SFP design temperature is exceeded.

Review of this modification under 10CFR50.59 was required because the addition of isolation capability to the SFP tell-tale leak-off lines constituted a change to the facility as described in the UFSAR. This modification provided additional protection from a loss of SFP inventory should SFP temperature exceed its maximum design value following a loss of cooling and cause failure of the SFP liner. Therefore, this change would not increase the probability or consequences of an accident previously analyzed. Additionally, this change did not increase the probability or consequences of a malfunction of equipment important to safety. This change would not create any new accidents or malfunctions since no new failure modes were introduced. In addition the Technical Specification Bases were not affected and no changes to the Technical Specifications were required.

• \ ,<

Temporary Modifications - Summary of Safety Evaluations Temporary Modification 99-014, Operation With Two Reactor Coolant System Hot Leg Resistance Temperature Detectors (RTD) In The 13 Reactor Coolant Loop This temporary modification removed the 13 Reactor Coolant hot leg RTD TE431A-2 from service in response to a steadily decreasing output. The two remaining hot leg RTDs will be used to provide indication, control, and protective functions until plant conditions permit the cause of the decreasing output to be determined and corrected.

Review of this temporary modification under 10CFR50.59 was required because the removal of RTD TE431A-2 from service constituted a change to the facility as described in the UFSAR. With RTD TE431A-2 isolated, the protective functions, indications, and controls utilizing signals supported by the hot leg temperature instrumentation will continue to function as described in the SAR. This change is supported by the current design and licensing basis. Therefore, this change would not increase the probability or consequences of an accident previously analyzed. Additionally, this chahge did not increase the probability or consequences of a malfunction of equipment important to safety. This change would not create any new accidents or malfunctions since no new failure modes were introduced. In addition the Technical Specification Bases were not affected and no changes to the Technical Specifications were required.

Procedures - Summary of Safety Evaluations There were no changes in this category implemented during August 1999.

UFSAR Change Notices - Summary of Safety Evaluations UFSAR Change Notice SCN99-048, Reactor Coolant System Chemical Degassing Operation During Shutdown This UFSAR change involved the addition of hydrogen peroxide to the Reactor Coolant System (RCS) following plant shutdown and cooldown for the purpose of chemically depleting RCS dissolved hydrogen inventory. Chemical degassing will occur only in operational Mode 5 and below.

Review of this temporary modification under 10CFR50.59 was required because chemical degassing of the RCS constituted a change to the facility as described in the UFSAR and constituted a change to procedures as described in the UFSAR. The evaluation concluded that this change would not increase the probability or consequences of an accident previously analyzed. Additionally, this change did not increase the probability or consequences of a malfunction of

equipment important to safety. This change would not create any new accidents or*malfur.ictions since no new failure modes were introduced. In addition the Technical Specification Bases were not affected and no changes to the Technical Specifications were required.

Deficiency Reports - Summary of Safety Evaluations There were no changes in this category implemented during August 1999.

Other - Summary of Safety Evaluations Salem Unit 1 Cycle 13 Reload Safety Evaluation for Extending Cycle Burn up This evaluation considered the extension of the Salem Unit 1 Cycle 13 reload burnup from 18000 to 19300 Megawatt Days/Metric Ton Uranium (MWD/MTU).

For Salem Unit 1, reload cores are analyzed for events contained in the Salem UFSAR, as well as limits given in the Technical Specifications (TS) and the Core Operating Limits Report (COLR). These analyses contain assumptions that involve parameters that are core design and burnup dependent. Therefore, a specific evaluation of safety parameters is required to confirm the validity of the existing safety analysis for the extended burn up.

This evaluation was reviewed under 50.59 because it the extension of the Salem Unit 1 Cycle 13 reload burnup constitutes a change to the facility as described in the UFSAR. The evaluation demonstrated that design limits would continue to be met. Therefore, the evaluation concluded that this change would not increase the probability or consequences of an accident previously analyzed.

Additionally, this change did not increase the probability or consequences of a malfunction of equipment important to safety. This change would not create any new accidents or malfunctions since no new failure modes were introduced. In addition the Technical Specification Bases were not affected and no changes to the Technical Specifications were required. '