ML18100A659

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Monthly Operating Rept for Sept 1993 for Salem Unit 1. W/931008 Ltr
ML18100A659
Person / Time
Site: Salem PSEG icon.png
Issue date: 09/30/1993
From: Heller R, Shedlock M
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9310200195
Download: ML18100A659 (13)
v • d • e


Text

PS~G*

Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station October 8, 1993 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 September 1993 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 Si~cer~ly~

~ager-Salem Operations RH:pc 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 People

~g~o2ooi95-9309-3~r* 1 95-2189 (10M) 12-89 R ADOCK 05000272 I,

.*,

ttvERAGE DAILY UNIT POWER L~L Docket No.: 50-272 Unit Name: Salem #1 Date: 10/08/93 Completed by: Mark Shedlock Telephone: 339-2122 Month September 1993 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 1049 17 1070 2 1080 18 1050 3 1067 19 1074 4 1109 20 1122 5 1089 21 1101 6 1085 22 1072 7 1085 23 1130 8 1085 24 1086 9 1084 25 1072 10 1069 26 1103 11 1118 27 1076 12 1100 28 1105 13 1068 29 1104 14 1069 30 1100 15 1070 31 16 1047 P. 8.1-7 R1

..

OPERATING DATA REPORT ~

Docket No: 50-272 Date: 10/08/93 Completed by: Mark Shedlock Telephone: 339-2122 Operating Status

1. Unit Name Salem No. 1 Notes
2. Reporting Period September 1993
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 NA 2
9. Power Level to Which Restricted, if any (Net MWe) N/A
10. Reasons for Restrictions, if any NONE This Month Year to Date Cumulative
12. Hours in Reporting Period 720 6551 142488
12. No. of Hrs. Rx. was Critical 720 5925.3 95107.3
13. Reactor Reserve Shutdown Hrs. 0 0 0
14. Hours Generator On-Line 720 5723.3 91863.7
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 2461305.6 18522156.0 290720764.4
17. Gross Elec. Energy Generated (MWH) 813540 6146790 96519780
18. Net Elec. Energy Gen. (MWH) 780873 5859566 91931225
19. Unit Service Factor 100. 87.4 64.5
20. Unit Availability Factor 100 87.4 64.5
21. Unit Capacity Factor (using MDC Net) 98.l 80.9 58.3
22. Unit Capacity Factor (using DER Net) 97.3 80.2 57.9
23. Unit Forced Outage Rate 0 12.6 21.0
24. Shutdowns scheduled over next 6 months (type, date and duration of each)

The Unit is presently in a refueling outage scheduled to last 72 da s.

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

NA 8-1-7.R2

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH SEPTEMBER 1993 DOCKET NO. :_5""'0'°"'"-=27'""2'"'="_ __

UNIT NAME: Salem #1 NONE DATE: 10/08/93 COMPLETED BY: Mark Shedlock TELEPHONE: 339-2'122 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 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 (NUREG-0161)

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

SAFETY RELATED MAINTE.CE DOCKE,NO: 50-272 MONTH: - SEPTEMBER 1993 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 WO NO UNIT EQUIPMENT IDENTIFICATION 930729237 1 1R5 SPENT FUEL PIT AREA RADIATION MONITOR FAILURE DESCRIPTION: 1R5 MONITOR FAILED - INVESTIGATE 930819068 1 VALVE 12SW264 FAILURE DESCRIPTION: LEAK UPSTREAM AT P3 CONNECTION -

REPAIR OR REPLACE 930913131 1 VALVE 11SW101 FAILURE DESCRIPTION: FLANGE ERODED ON VALVE - REPLACE 930924243 1 VALVE 1CV146 FAILURE DESCRIPTION: BORON ROCKED UP - REWORK AS REQUIRED 930927200 1 1R12A PLANT VENT GAS MONITOR FAILURE DESCRIPTION: MONITOR SPIKING AND ERRATIC -

TROUBLESHOOT

10CFR5*~. 59 EVALUATION,_ DOCKE,NO: 50-272 MONTH: - SEPTEMBER 1993 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations lOCFRS0.59. The Station Operations Review Committee has reviewed and concurs with these evaluations.

ITEM

SUMMARY

A. Design Change Packages (DCPs) lEC-3229 Pkg 1 "Fuel Transfer Tube Blind Flange Replacement/Bolt Reduction" - The proposed modification will replace the existing 20" fuel transfer blind flange at Salem Unit 1. The existing flange which uses two (2) flexitallic gaskets and requires twenty bolts to obtain a proper seal, will be replaced with a new blind flange utilizing a new seal design which only requires four bolts to obtain a proper seal. The new blind flange utilizes a Westinghouse design which replaces the flexitaliic gaskets with a rubber Quad ring seal design. Design calculations performed by Westinghouse show that a minimum of four (4) bolts can be used with the quad ring seals while still generating sufficient preload to compress the seals and resist hydraulic and seismic forces.

The proposed change to the fuel transfer tube blind flange is being implemented as a maintenance/ radiation dose reduction enhancement (ALARA). The margin of safety of the fuel transfer tube blind flange and bolting is defined, in part, by the requirements imposed by the ASME code. The design requirements have been shown to be met. The containment isolation function is maintained. The margin of safety is therefore, unaffected. (SORC 93-082) lEC-3267 Pkg 1 "MFOST Filter Secondary Containment" - This DCP would install a secondary containment beneath the fuel oil transfer piping and filter. The reason for this DCP is to meet DPCC Requirements for secondary containments around/under equipment for hazardous materials. Components of the proposed change are not discussed nor does it impact any component discussed in the basis for any Technical Specification. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-083)

10CFR'5*~. 59 EVALUATION,_ DOCKE,NO:

UNIT NAME:

50-272 SALEM 1 MONTH: - SEPTEMBER 1993 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

lEC-3243 Pkg 1 "SGFP/Condensate Pump/Feedwater Heaters: System Design Press Modifications" Rev. 1 - The purpose of this revision is to revise the 10CFR50.59 evaluation to address a change to UFSAR Table 10.4-1. This change will provide setpoint corrections for the 11, 12 and 13 condensate pump recirculation flow control loops to ensure a maximum sustained nominal pressure of 625 psig during cold startup. This change will also provide setpoint corrections for the f eedwater heaters 16A, 16B and 16C thermal relief valves 11, 12 and 13HV22. This change will develop a two-out-of-two high pressure trip signal by installing redundant transmitters on the discharge piping of each f eedwater pump to provide the required protection while minimizing the potential for a spurious trip. The existing feed pump trip circuitry will be modified to allow the feed pump to trip on a high feed pump discharge pressure at all power levels. The existing L&N power supply, bistable and transmitter from each of the #11 and

  1. 12 SGFP analog discharge pressure transmitter-circuits, as well as the existing Keylock override switches will be deleted to implement the new single channel instrument loop trip circuitry noted above. The only Technical Specification which involves SGFP circuitry is Technical Specification 3/4.3.2.b (Table 3.3-3, 5.a & 8.f), feedwater isolation and a trip of main feedwater pumps starting the motor driven Auxiliary Feedwater pump. This modification does not alter this function in any way. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-085) lEC-3208 Pkg 7 "Salem Fire Dampers Upgrade" Rev. 1 - The design scope for this package includes the addition of a new fire damper and modification on an existing fire dampers in the Control Area Air Conditioning System (CAACS). The function, basic configuration and operation of the system will not be altered and the codes, standards, qualification and design

10CF~50.59

.9 EVALUATIONS DOCKE,NO: 50-272 MONTH: SEPTEMBER 1993 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

criteria of the original system will apply. These modifications are intended to meet PSE&G's commitment to the NRC and compliance with the requirements of 10CFR50, Appendix R. The basis for the Technical Specifications does not address the fire protection features. The Technical Specifications discuss the surveillance requirements for the Control Room Emergency Air Conditioning System (EACS). Since replaced or modified fire dampers associated with the CAAC system perform no safety related functions; these modifications do not affect the safety related system which is required for safe shutdown. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-086) lEC-3191 Pkg 1 "Auxiliary Feedwater System 1AF128 Relief Valve Installation" - The proposed change entails the installation of a new pressure relief valve (1AF128) in the turbine driven auxiliary feed (TAFP) water pump continuous recirculation line.

This change will re-establish overpressurization protection for the No. 13 Aux. Feedwater pump turbine bearing coolers. All work will take place in the TAFP enclosure located on El. 84' in the auxiliary building. The scope of work for this design change, involves removing a spool piece and blind flange which were temporarily installed in place of relief valve 1AF128 which was removed by DCP lEC-3042. The new relief valve being installed is the same type (Crosby) as the one removed. The face to face dimensions are exactly the same, therefore, modifications to the existing piping system will not be required. In addition to installation of the new pressure relief valve, the valve stem retainer currently installed on valve 1AF103 (maintenance isolation valve), which was installed as part of DCP lEC-3042 to prevent inadvertent closure and potential overpressurization of the turbine jacket bearing coolers, will be removed and the valve locked in the open position. This modification is being made to reestablish overpressurization protection

10CFR50.59 EVALUATION~ DOCKE~NO:

UNIT NAME:

50-272 SALEM 1 MONTH: - SEPTEMBER 1993 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

for No. 13 Auxiliary Feedwater Pump Turbine Bearing Jacket Coolers located in the continuous recirculation line. The Woodward Governor/Turbine Bearing Jacket Coolers are required for the operation of No. 13 TAFP, which is a technical specification component. The installation of pressure relief valve 1AF128 will provide overpressure protection for these components, thus precluding any potential component failure due to overpressurization. There are no adversely impacted technical specifications and safety margins are not reduced. Thus, the margin of safety as defined in the basis for the Technical Specifications will not be changed. (SORC 93-086) lEC-3225 Pkg 1 "RG 1.97 EQ Valve Limit Switch and Solenoid *I Modifications" Rev. 1 - This design change package modifies the existing power, wiring and electrical protective devices of Environmentally Qualified (EQ) solenoids and valve position limit switches to provide electrical separation from Non-Environmentally Qualified (Non-EQ) solenoids and valve position limit switches. This design change package was developed to ensure that Regulatory Guide 1.97, Category 1, Type B Variable, valve position indication for the Containment Isolation Valves listed previously, possess proper electrical separation between EQ related components and Non-EQ related components.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-086)

B. Procedures and Revisions NC.NA-AP.ZZ-0003(Q) "Document Control Program" Rev. 2 - This procedure was revised to comply with the requirements of NC.NA-AP.ZZ~OOOl(Q), Nuclear Department Procedure System, NC.NA-AP.ZZ-0032(Q),

Preparation, Review and Approval of Procedures and

1cicFR50.59 EVALUATION- DOCKE.O: 50-272 MONTH: - SEPTEMBER 1993 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

NC.NA-AS.ZZ-OOOl(Q) Nuclear Department Procedure Format and Content Standard. This revision meets the biennial review requirements specified in NC.NA-AP.ZZ-0032(Q), Preparation, Review and Approval of Procedures. Due to the administrative nature of the procedure it does not create the possibility of an accident or malfunction of a different type than any previously evaluated. The Technical Specifications do not address the document control program. Therefore, this revision cannot reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-082)

NC.NA-AP.ZZ-OOll(Q) "Records Management Program" Rev. 2 - This procedure was revised to comply with the requirements of NC.NA-AP.ZZ-OOOl(Q). Nuclear Department Procedure system, NC.NA-AP.ZZ-0032(Q),

Preparation, Review and Approval of Procedures and NC.NA-AP.ZZ-OOOl(Q) Nuclear Department Procedure Format.and Content Standard. This revision meets the biennial review requirements specified in NC.NA-AP.ZZ-0032(Q), Preparation, Review and Approval of Procedures. The Technical Specifications discuss the records to be retained in the Administrative Section and do not use that information for the basis of any Technical Specification. Therefore, this revision does not reduce the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-082)

NC.NA~AP.ZZ-0036(Q) "Control of Information System and Telecommunications Services" Rev. 2 - This revision: 1.) Incorporated the Help Desk for informal requests for services related to information systems and telecommunications services; 2.) Removed sections on excluded items and moved the exclusion to the scope section.

Also removed training computers from the scope; 3.) Incorporated the request of telecommunications

10CFR50.59 EVALUATION- DOCKE-0: 50-272 MONTH: - SEPTEMBER 1993 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

services; 4.) Added requirements to use the Nuclear Department Resource Allocation Program (NDRAP) process for application requests that meet the requirements of the NDRAP process: 5.) Added responsibility for the User to arrange for a disaster recovery method for software systems under their control, and 6.) Incorporated the home computer policy into this procedure from the memo NMS-90-0041, dated 08/10/90. The Technical Specifications do not address information or telecommunications systems or services.

Therefore, there is no reduction in the margin of safety for any Technical Specification.

(SORC 93-082)

NC.NA-AP.ZZ-0061(Q) "Significant Event Response Team Management" - The changes associated with this revision primarily serve to simplify the previously approved procedure and include the following: a.) Modified the purpose of the procedure to provide guidance*

for SERT Managers for completing independent assessments. All procedure sections reflect the new purpose; b.) The intent of the revision is to make the procedure a tool for the intended user.

Since the intended user is an experienced manager, the level of detail has been reduced; c.)

Identified a SERT position of Information Coordinator. Previously, this position was limited to a member of the Technical Department; d.) Added attachment listing SERT action items to be considered. This list provides the members with the necessary guidance without unnecessary explanations. This procedure controls the SERT process. It does not direct the operation of equipment, either safety related or non safety related and does not affect any equipment or process controlled by Technical Specifications; therefore it cannot reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-086)

\.

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 SEPTEMBER 1993 SALEM UNIT NO. 1 The Unit began the period operating at full power and continued to operate at essentially full power throughout remainder of the period.

RiFUELING INFORMATION MONTH: - SEPTEMBER 1993 e DOCKE.O: 50-272 UNIT NAME: SALEM 1 DATE: OCTOBER 8, 1993 COMPLETED BY: R. HELLER TELEPHONE: {609)339-5162 MONTH SEPTEMBER 1993

1. Refueling information has changed from last month:

YES X NO

2. Scheduled date for next refueling: OCTOBER 2, 1993
3. Scheduled date for restart following refueling: DECEMBER 13, 1993
4. a) Will Technical Specification changes or other license amendments be required?:

YES NO X NOT DETERMINED TO DATE b) Has the reload fuel design been reviewed by the station Operating Review Committee?:

YES NO x If no, when is it scheduled?: NOVEMBER

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 732
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
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