ML18096A070

Monthly Operating Rept for May 1991 for Salem Unit 1. W/910615 Ltr

ML18096A070
Person / Time
Site:	Salem
Issue date:	05/31/1991
From:	Shedlock M Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9106210078
Download: ML18096A070 (11)

Text

e OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station June 15, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

MONTHLY OPERATING REPORT SALEM NO. 1 DOCKET NO. 50-272 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 1991 are being sent to you.

Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions Safety Related Maintenance 10CFR50.59 Evaluations Operating Summary Refueling Information

?4:i4:_

General Manager -

Salem Operations RH:pc cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-l-7.R4

'-. T:H-o:l+/-rj_e'[gy Peoole __ _

r.., _,_* tt v ' 9* :L* n ~ --=* 1 (><">?*=*

- '-'L .. .. '-' 9105::::1.

PDR ADOCI< 0!5000272 95-2189 (10M) 12-89 F\: PDR

~ERAGE DAILY UNIT POWER L~L Docket No.: 50-272 Unit Name: Salem #1 Date: 6/10/91 Completed by: Mark Shedlock Telephone: 339-2122 Month May 1991 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 1085 17 1093 2 1100 18 1077 3 1097 19 1077 4 1115 20 1040 5 1121 21 1065 6 1100 22 1051 7 1103 23 1084 8 1099 24 1100 9 1109 25 1099 10 1112 26 1085 11 1117 27 1094 12 1117 28 1108 13 1094 29 1077 14 1094 30 1118 15 1107 31 1094 16 1093 P. 8.1-7 R1

. . OPERATING DATA REPORT e Docket No: 50-272 Date: 6/10/91 Completed by: Mark Shedlock Telephone: 339-2122 Operating Status

1. Unit Name Salem No. 1 Notes

2. Reporting Period May 1991

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.~-=N:..i...:.A~~~~~~~~~~~~~~~~~~~~~~~~

2

9. Power Level to Which Restricted, if any (Net MWe) N/A

10. Reasons for Restrictions, if any ~~~~~~~N=-=+-=A=----~~~~~~~~~~~

This Month Year to Date Cumulative

12. Hours in Reporting Period 744 3623 122016

12. No. of Hrs. Rx. was Critical 744 1886.1 78849.6

13. Reactor Reserve Shutdown Hrs. 0 0 0

14. Hours Generator on-Line 744 1799.0 76366.5

15. Unit Reserve Shutdown Hours 0 0 0

16. Gross Thermal Energy Generated (MWH) 2536202.4 5964273.6 240026658.8

17. Gross Elec. Energy Generated (MWiI) 848170 1978070 79690710

18. Net Elec. Energy Gen. (MWH) 814143 1881071 75849743

19. Unit Service Factor 100 49.7 62.6

20. Unit Availability Factor 100 49.7 62.6

21. Unit Capacity Factor (using MDC Net) 98.9 46.9 56.2

22. Unit Capacity Factor (using DER Net) 98.1 46.6 55.8

23. Unit Forced outage Rate 0 0 22.1

24. Shutdowns scheduled over next 6 months (type, date and duration of each)

NONE

25. If shutdown at end of Report Period, Estimated Date of Startup:

NA 8-l-7.R2

UNIT SHUTDOYN AND POYER REDUCTIONS REPORT MONTH MAY 1991 DOCKET NO. 50-272 UNIT NAME Salem #1 DATE 6/10/91 COMPLETED BY Mark Shedlock TELEPHONE 339-2122 METHOD OF SHUTTING LICENSE DURATION DOYN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE 1 (HOURS) REASON 2 REACTOR REPORT # CODE 4 CODE 6 TO PREVENT RECURRENCE 1 2 3 4 5 F: Forced Reason Method: Exhibit G - Instructions Exhibit 1 - Same S: Scheduled A-Equipment Failure (explain) 1-Manual for Preparation of Data Source B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling 3-Automatic Scram Event Report CLER) File D-Requlatory Restriction 4-Continuation of CNUREG-0161)

E-Operator Training & License Examination Previous Outage F-Administrative 5-Load Reduction G-Operational Error (Explain) 9-0ther H-Other (Explain)

SAFETY RELATED MAINTENANCE DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: JUNE 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 WO NO UNIT EQUIPMENT IDENTIFICATION 900419103 1 G.I.C. UNITS FAILURE DESCRIPTION: TROUBLESHOOT GEOMAGNETIC INDUCTION COUNTERS 901130209 1 VALVE 11MS178 FAILURE DESCRIPTION: VALVE IS STEAM CUT - REPLACE VALVE 910109164 1 SPOOL 1-SW-P-2010 FAILURE DESCRIPTION: SPOOL ERODED AT FLANGE - REPLACE 910109165 1 SPOOL 1-SW-2053B FAILURE DESCRIPTION: SPOOL ERODED AT FLANGE - REPAIR 910317077 1 SPOOL 1-SW-2344NA FAILURE DESCRIPTION: REPAIR SPOOL AS PER DR SMD 91-195 910330092 1 SPOOL RD-1-SW-58 FAILURE DESCRIPTION: SPOOL/FLANGE LEAK - REPAIR 910402176 1 VALVE 12SW253 FAILURE DESCRIPTION: VALVE IS PLUGGED - REPLACE 910418120 1 11 BAT PUMP FAILURE DESCRIPTION: 11 BORIC ACID TRANSFER PUMP MECH.

SEAL IS LEAKING - REPLACE 910424181 1 SPOOL 1-SW-36 FAILURE DESCRIPTION: REPAIR THROUGH WALL LEAK -

REPLACE DEGRADED SPOOL 'SECTION 910507065 1 14 S/G LEVEL INDICATION FAILURE DESCRIPTION: WIDE RANGE LEVEL/PRESS RECORDER ERRATIC - TROUBLESHOOT

r

'

/

-

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: JUNE 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (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 these evaluations and agrees with their conclusions that these items do not present Unreviewed Safety Questions (USQs).

ITEM

SUMMARY

A. Design Change Packages lEC-3089 Pkgs. 2 & 4 "Reactor Coolant Pump Oil Lift Pump & Motor Assembly Enclosure Modifications" - Package 2 proposes the modification of the existing No. 12 RCP oil lift pump & motor assembly enclosure to close up some gaps that exist in the enclosure.

This will ensure the oil lift pump motor is completely isolated from the enclosure.

Additional mounting bolt holes will be drilled into the RCP motor flange to adequately support the modified enclosure. There is no change in either the RCP system operation or testing procedure.

Package 4 proposes the modification of the existing No. 14 RCP oil lift pump & motor assembly enclosure to accommodate the new larger frame size RCP oil lift pump motor. This modification is restricted to structural changes involving the re-design of the left panel of the motor & pump enclosure to clear the rear of the new TEFC motor. Additional mounting bolt holes will be drilled into the RCP motor flange to adequately support the modified enclosure.

There is no change in either the RCP system operation or testing procedure. (SORC 91-034) lEC-3065 Pkg. 1 "Narrow Range RTD Conax Connections" - As a result of this design change, the connections between the Conax connectors and the Narrow Range RTDs will be made by butt slicing the leads together, and then soldering the butt spliced connections with a high temperature solder. Shrink tubing will be placed over the soldered joints to protect against possible shorting, and the terminal posts inside the RTD heads will be removed. All special considerations associated with the use of high temperature solder have been incorporated into the DCP. The proposed modifications meet the requirements of the original design specifications. ( SORC 91-030)

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: JUNE 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (Cont'd)

ITEM

SUMMARY

SE # NFU 91-165 "Salem Unit 1 Cycle 10 Final Loading Pattern (Redesigned) Reload Safety Evaluation (RSE) and NRC Notification" - This revised RSE presents a re-evaluation for Salem Unit 1, Cycle 10, which demonstrates that the conditions described in Section 1.3 for the core reload will not adversely affect the safety of the plant. This evaluation was accomplished using the metho~ology described in WCAP-9273-NP-A, "Westinghouse Reload Safety Eval~ation Methodology". The Salem Unit 1 Cycle 10 design has been re-evaluated due to a revised core loading pattern, caused by damage to two Region llA, one Region SB and one region 5A fuel assemblies (observed at the Cycle 9/10 refueling) . The damaged assemblies are being replaced by three Region 10 and one Region 6 fuel assemblies. In addition, some fuel assembly shuffles were performed to enhance the overall power distribution. This re-evaluation assures that all core design and safety limits are satisfied with the actual Cycle 9 burnup of 16520 MWD/MTU and the revised Cycle 10 loading pattern. All accident analyses have been reviewed and found to be satisfactory. The current FSAR is still applicable to the redesigned core as the core meets the requirements of the Chapter 15 bounding analysis. (SORC 91-032)

I I

I I

I

" ..

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 MAY 1991 SALEM UNIT NO. 1 The Unit began the period operating at full power, and, with the exception of brief load reductions for condenser waterbox cleaning, continued to operate at full power throughout the period.

REFUELING INFORMATION DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: JUNE 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 MONTH MAY 1991

1. Refueling information has changed from last month:

YES NO X

2. Scheduled date for next refueling: APRIL 18, 1992

3. Scheduled date for restart following refueling: JUNE 12, 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 NO x 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 588

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: September 2001 8-1-7.R4

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: JUNE 10, 1991 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (Cont'd)

ITEM

SUMMARY

lEC-3049 Pkg. 2 "1Rl3A and 1Rl3B Warning and Alarm Setpoints" -

The purpose of this design change is to implement the new Warning and Alarm setpoints for 1Rl3A and 1Rl3B Containment Cooler Radiation Monitors. Previously it had been noted there was a difference in the implemented setpoints for the 1R13A and 1R13B radiation monitors.

This DCP will modify the setpoints to common setpoints. The differences in the setpoints resulted from a slight change in the calculation methodology. The original setpoints were verified to be conservative. (SORC 91-047)

B. Deficiency Reports (DRs) Use-As-Is Dispositions DR # SMT 91-190A "11 Steam Generator" - The tube located at Row 1 column 2 is ruptured. the indication is approximately 1 3/8" long. The rupture is located at the area where the tube lane blocking device was previously located. This tube is plugged with explosive plugs. The existing explosive plugs in the tube do not show any evidence of leakage. (SORC 91-033)

DR # ISI-91-271 "14 Steam Generator" - Fosar examination of the secondary side of 14 S/G revealed 1 piece of thin wire, less than 1/16" diameter, length unknown, located at column 86 row 28. Fosar attempted to remove the wire but could not. The wire is stuck in the tube bundle. The resolution of this DR states the thin wire in 14 SIG may be left in place until the lRlO outage, provided the following activities are performed:

1) The tubes in contact with the identified object are eddy current tested during the lRlO outage in order to determine tube integrity and wear rate, and 2) A follow up work order is initiated to attempt removal of the wire during the lRlO outage. If the wire cannot be removed at that time, an additional disposition will be required. Westinghouse had previously performed a similar evaluation for Salem Unit 2 which showed a Use-As-Is disposition to be acceptable. That case involved a similar size wire. As in the case of Unit 2, the wire in 14 S/G cannot currently be retrieved. The only potential problem is a possible tube failure caused by wear. However, the most recent eddy current examinations of the affected tubes indicate no wear has occurred. (SORC 91-049)

lOCFRS0.59 EVALUATIONS DOCKET NO: 50-272 MONTH: - MAY 1991 UNIT NAME: SALEM 1 DATE: . JUNE 10 , 19 91 COMPLETED BY: J. FEST TELEPHONE: (609)339-2904 (Cont'd)

ITEM

SUMMARY

c. Procedures and Revisions Sl.OP-PT-SW-0017(Q) "No. 12 Component Cooling Heat Exchanger Heat Transfer Performance Data Collection", Rev. 0 -

This procedure provides instruction necessary to collect heat transfer data for No. 12 Component Cooling Heat Exchanger. The data collected shall be used in a performance evaluation to meet the intent of NRC Generic Letter. 89-13.

All temporary instruments will be removed at the completion of this test and the permanent instruments that were removed will be reinstalled. (SORC 91-047)

AIT&QP "Artificial Island Security Training &

Qualification Plan", Rev. 2 - This revision to the procedure is primarily to comply with New Jersey state law N.J.S.A. 2C:39-6J regarding firearms qualification and requalification for law enforcement and private security personnel.

This revision brings the procedure requirements in line with police requirements. (SORC 91-047)

D. Safety Evaluations (S/E)

SE # SECL-91-148 "Steam Generator Tube Explosive Plug Safety Evaluation", Rev. 1 - This evaluation is written to assess the* safety impact of plant operation of Salem Unit 1 with explosive plugs being returned to service. The structural adequacy of f ishmouthed tubes RlC2 and RlC3 of Steam Generator 14 during future operation, the potential bulging and fishmouthing of other explosively plugged tubes, and the probability and consequences of potential primary to secondary leakage from an explosively plugged tube for all plant conditions, are evaluated.

The effect of localized degradation on plugged tube burst capability is also assessed. This evaluation is limited in scope to tubes that have been removed from service utilizing explosive plugs. The hydraulic ratcheting effect that is postulated to have caused the burst of tubes R1C2 and R1C3 has neither been observed nor is expected to occur in a tube removed from service utilizing a Westinghouse mechanically expanded plug. Leakage monitoring will be continued. (SORC 91-033)