ML18101A632

Monthly Operating Rept for Mar 1995 for Salem 1.W/950413 Ltr

ML18101A632
Person / Time
Site:	Salem
Issue date:	03/31/1995
From:	Heller R, Phillips R, Summers J Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9504180127
Download: ML18101A632 (11)

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Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit April-13, 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk MONTHLY OPERATING ~PORT SALEMNO. 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 March are being sent to you.

~=-

John C. Summers General Manager -

Salem Operations RH:vls Enclosures C Mr. Thomas T. Martin Regional Administrator USNRC, Region I 631 Park Avenue King of Prussia, PA 19046 8-l-7.R4 The power)1:_in_vom:_b'.lnds____ _

r 9504100127 950331 i PDR ADOCK 05000272 95-2168 REV. 6/94

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ta.ERAGE DAILY UNIT POWER LE~

Docket No.: 50-272 Unit Name: Salem #1 Date: 04/10/95 Completed by: Robert Phillips Telephone: 339-2735 Month March 1995 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 0 17 1128 2 254 18 972 3 462 19 1009 4 334 20 1122 5 472 21 1045 6 472 22 1051 7 853 23 1048 8 1106 24 1079 9 1122 25 1022 10 1120 26 1034 11 1104 27 1046 12 1124 28 1051 13 1126 29 1059 14 1126 30 1067 15 1092 31 1066 16 1109 P. 8.1-7 Rl

e OPERATING DATA REPORT e

Docket No: 50-272 Date: 04/10/95 Completed by: Robert Phillips Telephone: 339-2735 Operating Status

1. Unit Name Salem No. 1 Notes

2. Repo,rting Period MARCH 1995

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~*~A=---~~~~~~~~~~~~~~~~~~~~~~~

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

10. Reasons for Restrictions, if any NA This Month Year to Date Cumulative

12. Hours in Reporting Period 744 2160 155617

12. No. of Hrs. Rx. was Critical 744 1554.6 103274.2

13. Reactor Reserve Shutdown Hrs. 0 0 0

14. Hours Generator On-Line 719 1527.1 99283.2

15. Unit Reserve Shutdown Hours 0 0 0

16. Gross Thermal Energy Generated (MWH) 2152334.4 4855963.2 314907866.0

17. Gross Elec. Energy Generated (MWH) 719610 1636370 104247520

18. Net Elec. Energy Gen. (MWH) 687443 1553273 99235437

19. Unit Service Factor 96.6 70.7 63.8

20. Unit Availability Factor 96.6 70.7 63.8

21. Unit Capacity Factor (using MDC Net) 83.5 65.0 57.7

22. Unit Capacity Factor (using DER Net) 82.9 64.5 57.2

23. Unit Forced Outage Rate 3.4 29.3 21.4

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

We are scheduled to start a 60 day refueling outage on 9/9/95.

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

NA 8-1-7.R2

UNIT SHUTDOWN AND POWER REDUCTIONS DOCKET NO: 50-172 REPORT MONTH MARCH 1995 UNIT NAME: Salem# 1 DATE: 4-10-95 COMPLETED BY: Robert Phillips TELEPHONE: 609-3'.39-2735 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE 1 (HOURS) REASON 2 REACTOR REPORT # CODE 4 CODE 5 TO PREVENT RECURRENCE 2373 3-1-95 F 25 A 4 cc INSTRU NUCLEAR MAIN STEAM SAFETY/CONTROLS 2407 3-4-95 F 52.5 D 5 ------------ zz zzzzzz NUCLEAR OTHER FUEL LIMITS 2489 3-18-95 F 14.6. F 5 ------------ zz zzzzzz MARSH FIRE UNDER LINE 5015 91 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)

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R.HELLER TELEPHONE: 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.

1. Design Change Packages (DCP) lEA-1084, Pkg. 1 "Update Documents to Reflect Deletion of N 2 Line to PWST" Rev. 0 - The changes to the following documents will reflect the original intent ofDCR MD-0010; Drawing Nos. 205240 (P&ID), 205230 (P&ID), 218253 (Mech Arrgmnt), 207478 (Mech Arrgmnt), 230527 (Mech Pipe Detail), and 226143 (Hanger Details). The actual document changes will involve deleting the "hard piping" run from valve 1NT900 to valve 1WR902. Hose connections are available in lieu of hard piping to serve as a path for nitrogen to enter the Primary Water Storage Tank (PWST) for purging. The as-built condition of nitrogen purge to the PWST does not affect the operability or maintainability of either the Bulk Nitrogen Packages or the PWST. No system or equipment serving a safety function is affected by this as-built DCP. Therefore, this DCP does not reduce the margin of safety as defined in the bases for any Technical Specifications. (SORC 95-026) lEC-3405, Pkg. 1 "Unit 1 Waste Gas Analyzer Panel Backfit" Rev. 0 - The purpose of this design change is to perform the internal and external modifications at panel 110-1 that are required to make the panel identical to panel 110-2. There are no new credible failure modes associated with this change. The portions of the Waste Gas System modified by this DCP are all non-safety related. All equipment added, replaced or modified at panel 110-1 will perform identical functions as the existing, with improved operation. These modifications improve the existing waste gas analyzer panel for 0 2 and H 2 monitoring of in-service waste decay tank and source tank cover gas. The changes will not increase the risk of

IOCFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R.HELLER TELEPHONE: 609-339-5162 explosivity and do not have an effect on monitoring of gaseous effluents. Since no safety related system is affected by this modification, the margin of safety is not reduced.

(SORC 95-026) lEC-3345, Pkg. 1 "Turbine Runback/SGFP Control Circuit Modification" Rev. 0 - This DCP installs circuitry which will initiate automatic runback of the main turbine, open condensate polishing by-pass valves l 1CVN108, 12CN108 and 13CN108 and open feedwater heater by-pass valve 1CN47 upon trip of a steam generator feedpump (SGFP). The runback will occur and the valves will open above a turbine power level approx. 60-70%, without operator intervention, to prevent a reactor trip. All circuitry installed by the automatic turbine runback portion of this DCP is non-safety related with the exception of the steam generator feed pump (SGFP) auxiliary latching relay (5X) and fail safe auxiliary relays (5X-2 and 5X-3). Credible failure modes associated with both the safety and non-safety related portions of the circuitry installed by this DCP have been addressed.

Failures which result in a complete turbine runback to 0%

power and/or runback at a rate of200% /minute are bounded in the existing SAR (i.e., Loss of electrical Load and/or Turbine Trip), and present no hazard to the integrity of the RCS or the Main Steam system. Failures which result in automatic runback not being initiated when a SGFP trips do not inhibit Operator manual actuation of turbine runback and opening the valves. Automatic runback is not a requirement. There is no reduction in the margin of safety as defined in the bases for any Technical Specification.

(SORC 95-029) lEC-3394, Pkg. 1 "Feedwater Flow Measurement System" Rev. 0 - This DCP replaces the existing feedwater flow metering nozzles 1FE510, 1FE520, 1FE530 and 1FE540 and adds new leading edge ultrasonic meters (LEFMs) lFl 7489, lFI 7490, lFl 7491 and lFl 7492, in the pump discharge.

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R. HELLER TELEPHONE: 609-339-5162 Replacement and added flow metering devices will be installed with flanged connections for removal and inspection (current configuration is welded pipe connections except for 1FE540, line 14). The existing indicators, 1FL8924I, 1FL8925 l, 1FL8926I and 1FL8927I, will also be replaced with new indicators to accommodate the increased differential pressures across the new nozzles.

Constants used in the P250, SPDS and ERDS will be changed to reflect the new flow nozzles. There is a Reactor Trip system signal associated with Steam/Feedwater flow Mismatch and Low Steam Generator Water Level. The feedwater flow control circuitry associated with the replacement of the flow metering nozzles was not affected by the nozzle replacement. The reactor trip setpoints were not altered by this change. The proposed modification does not reduce the margin of safety as defined in the bases for any Technical Specifications. Other than the reactor trip signal discussed above, the Feedwater System is not addressed in the Technical Specifications or their bases. As such, there are no requirements on the Feedwater System imposed by the Technical Specifications and no reduction in the margin of safety. (SORC 95-029) lEA-1089, Pkg. 1 "As-Built Documentation of A-36 Plate Used in lA-SGBLS-29" Rev. 0 - The purpose of this DCP is to assess and document differences between the as-built configuration and existing design documentation. Drawing 238963, Sheet 1, Rev. 7 is being updated to indicate encapsulation sleeves are considered safety related. Drawing 238963, Sheet lOA, Rev. 0 is being revised to indicate that A-36 plate material is acceptable for use. UFSAR Section 3.6.5.11 is being revised to clarify NDE requirements for encapsulation sleeves as described in ANSI B3 l. 7. Encapsulation sleeves are not addressed in the Technical Specification. The intended function and design of encapsulation sleeve lA-SGBLS-29 are not affected. Therefore, the changes do

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R. HELLER TELEPHONE: 609-339-5162 (Cont'd) not reduce the margin of safety as defined in the bases for any Technical Specifications. (SORC 95-030) lEC-3368, Pkg. 1. "No. 11, 12, 13 and 14 Reactor Coolant Pump Thermal Barrier Pressure Tap Modifications/CV251 and 252 Valve Elimination" Rev. 0 - The proposed modification entails a minor change in configuration to unused flanged pressure tap connections (two (2) per pump) attached to No. 11, 12, 13 and 14 reactor coolant pumps. There is prior experience with leakage at one of these pressure tap flange connections, which has resulted in a delayed unit startup and could potentially result in a unit shutdown. The intent of this modification is to simplify the present design by eliminating the pressure tap flange connections and non-functional root valves CV251 & 252 and welding on a pipe cap. This will in effect improve the long term reliability of the connections by eliminating a potential source of future leakage. The proposed modification represents only a minor configuration change to enhance the long term reliability of the subject pipe connections. There is no effect on the margin of safety as defined in the bases for any Technical Specifications. (SORC 95-032) lEC-3342, Pkg. 1 "Circulating Water Bearing Lubrication Screen Wash Setpoint & System Changes" Rev. 0 - The plant modifications to be accomplished under this DCP include:

1) the recalibration of header pressures and flows of the Circulating Bearing Lubrication system and the Screen Wash system, 2) Lubrication Pump and Screen Wash Pump strainer control changed from timed to continuous, 3) replacement of the lubrication water and screen wash pressure gages, and, 4) replacement of control valves with manual valves to simplify the system. The affected components do not change in regard to their function. No new failure modes will be introduced by these modifications.

10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R. HELLER TELEPHONE: 609-339-5162 There is no reduction in the margin of safety as defined in the basis for any Technical; Specification. (SORC 95-032

2. Safety Evaluations (S/E)

S-1-SW-MSE-0847 "Salem Unit 1 Service Water Pump Fastener Material" Rev.

0 - This SE addressees concerns pertaining to fasteners used in the column-to-discharge head connections on Salem Unit 1 Service Water (SW) pumps. The evaluation is based on the conditions described in Memo of 3/24/95 from J. Barnes to J. Lin. The conditions became apparent through Problem Report No. 950119152 which described the potential for Service Water pumps having stainless steel (SS) fasteners in place of the Monet fasteners originally supplied in the pumps. The fasteners have been found to be acceptable for short term use, as discussed in this evaluation. Therefore, the condition does not reduce the margin of safety as defined in the basis for any Technical Specification. Being that the design integrity of the two applicable SW pumps is maintained and the remaining four pumps have the correct fasteners, operability of the six SW pumps is not affected by this condition. (SORC 95-031)

REFUELING INFORMATION DOCKET NO: 50-272 MONTH: MARCH 1995 UNIT NAME: SALEM 1 DATE: 04/10/95 COMPLETED BY: R.HELLER TELEPHONE: 609-339-5162 MONTH: MARCH 1995 Refueling information has changed from last month: YES NO_X_

Scheduled date for next refueling: September 9, 1995 Scheduled date for restart following refueling: November 7. 1995

a. Will Technical Specification changes or other license amendments be required?

YES NO ___

NOT DETERMINED TO DATE _A_

b. Has the reload fuel design been reviewed by the Station Operating Review Committee?

YES NO_X_

If no, when is it scheduled? ~September 1995 Scheduled date(s) for submitting proposed licensing action: ---~NIA.____

Important licensing considerations associated with refueling:

Number of Fuel Assemblies:

a. Incore 193

b. In Spent Fuel Storage 732 Present licensed spent fuel storage capacity: _1632_ _

Future spent fuel storage capacity: _1632_ _

Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: September 2008 8-1-7.R4

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 MARCH 1995 SALEM UNIT NO. 1 The Unit began the period in Mode 1, Power Operation, and synchronized to the grid on 03/02/95, with power increasing. Power was increased to 48% to perform reactor physics testing and nuclear instrument calibrations. On 03/06/95, following completion of NI calibrations, a power escalation commenced. The Unit reached 100% reactor power on 03/08/95, and continued to operate at that level until 03/18/95. Power was reduced to 70% for a short period on 03/18/95, due to a meadow fire underneath the Keeney Line.

The line had been removed from service and required a load reduction due to system loading restrictions. The Unit returned to 100% power later the same morning. Power was reduced to 94% on 03/21/95 due to a leak in No. 1 lC Feedwater Heater. The Unit continued to operate at 94% power throughout the remainder of the period.