ML18096A637

Monthly Operating Rept for Mar 1992 for Salem Unit 1.W/

ML18096A637
Person / Time
Site:	Salem
Issue date:	03/31/1992
From:	Shedlock M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9204210185
Download: ML18096A637 (13)

Text

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OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station April 13, 1992 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.l.6, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating reports for the month of March 1992 are being sent to you.

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

~c~rely yours,

~'~

Salem Operations cc:

Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-l-7.R4 T~e i=_n~rgy_i:>e_ople 9204210185 920331

.. --l 8 PDR ADOCK 05000272 R

PDR 95*2189 (10M) 12-89

I AVERAGE DAILY UNIT POWER LEVEL Completed by:

Mark Shedlock Month March 1992 Day Average Daily Power Level (MWe-NET) 1 1117 2

1116 3

1120 4

1120 5

1121 6

1119 7

1117 8

1121 9

1112 10 1120 11 1089 12 1125 13 1125 14 1125 15 1119 16 1115 P. 8.1-7 Rl Day Docket No.:

50-272 Unit Name:

Salem #1 Date:

04/10/92 Telephone:

339-2122 Average Daily Power Level (MWe-NET) 17 1116 18 1130 19 1130 20 1118 21 1119 22 1119 23 1123 24 1112 25 1122 26 1116 27 1099 28 1126 29 1126 30 1108 31 1118

OPERATING DATA REPORT Docket No:

Date:

Completed by:

Mark Shedlock Telephone:

Operating Status

1.

Unit Name Salem No. 1 Notes

2.

Reporting Period March 1992

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)

Report, Give Reason NA

9.

Power Level to Which Restricted, if any (Net MWe)

10. Reasons for Restrictions, if any

11. Hours in Reporting Period

12. No. of Hrs. Rx. was Critical

13. Reactor Reserve Shutdown Hrs.

14. Hours Generator On-Line

15. Unit Reserve Shutdown Hours

16. Gross Thermal Energy Generated (MWH)

17. Gross Elec. Energy Generated (MWH)

18. Net Elec. Energy Gen. (MWH)

19. Unit Service Factor

20. Unit Availability Factor

21. Unit Capacity Factor (using MDC Net)

22. Unit Capacity Factor (using DER Net)

23. Unit Forced outage Rate This Month 744 744 0

744 0

2538681.6 865740 831920 100 100 101.1 100.3 0

NA Year to Date 2184 2184 0

2046.1 0

6964245.6 2335270 2238572 93.7 93.7 92.7 91.9 6.3 50-272 04/10/92 339-2122 since Last N/A Cumulative 129337 85784.3 0

83094.0 0

262461390.8 87139900 83015145 64.2 64.2 58.0 57.6

21. 0

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

Refueling outage scheduled to start 4-4-92 and last 73 days.

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

NA 8-l-7.R2

NO.

DATE 1

2 F:

Forced S:

Scheduled UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH MARCH 1992 DURATION TYPE1 (HOURS)

REASON2 Reason A-Equipment Failure (explain)

B-Maintenance or Test C-Refueling D-Requlatory Restriction METHOD OF SHUTTING DOWN REACTOR E-Operator Training & License Examination F-Administrative G-Operational Error (Explain)

H-Other (Explain) 3 LICENSE EVENT REPORT #

Method:

1-Manual 2-Manual scram SYSTEM CODE4 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther 4

DOCKET NO.

UNIT NAME DATE COMPLETED BY TELEPHONE 50-272 Salem #l 04/10/92 Mark Shedlock 339-2122 COMPONENT CAUSE AND CORRECTIVE ACTION CODE5 Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161)

TO PREVENT RECURRENCE 5

Exhibit 1 - Same Source

e e

SAFETY RELATED MAINTENANCE MONTH: -

MARCH 1992 DOCKET NO:

UNIT NAME:

50-272 SALEM 1 DATE:

COMPLETED BY:

TELEPHONE:

APRIL 10, 1992 J. FEST (609)339-2904

---~----------------------------------------------------------------------

WO NO UNIT 900615143 1

911104131 1

920113138 1

920129157 1

920228132 1

920317126 1

920323134 1

EQUIPMENT IDENTif ICATION RADIATION MONITOR 1R12A FAILURE DESCRIPTION:

FAULTY CHECK SOURCE -

INSPECT &

REPAIR 1C DIESEL GENERATOR FAILURE DESCRIPTION:

AIR START REGULATOR VALVE DID NOT FAIL OPEN -

INVESTIGATE RADIATION MONITOR 1R12A FAILURE DESCRIPTION:

DETECTOR CONNECTION HAS BARE CONDUCTOR -

REPAIR

1. 15 CONTAINMENT FAN COIL UNIT (CFCU)

FAILURE DESCRIPTION:

14 CFCU WATER HAMMER -

TROUBLESHOOT & REWORK 1C DIESEL GENERATOR FAILURE DESCRIPTION:

1C D/G SUSPECTED CLOG IN STARTING AIR -

INVESTIGATE & CORRECT 12 CHARGING PUMP ROOM COOLER FAILURE DESCRIPTION:

12 CHARGING PUMP ROOM COOLER DOES NOT WORK IN AUTOMATIC -

INVESTIGATE 15 SERVICE WATER PUMP FAILURE DESCRIPTION:

15 SW PUMP HIGH VIBRATION -

INVESTIGATE

10CFR50.59 EVALUATIONS MONTH: -

MARCH 1992 e

DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

50-272 SALEM 1 APRIL 10, 1992 J. FEST (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 and concurs with these evaluations.

ITEM

SUMMARY

A.

Design Change Packages (DCP}

DCP# lEC-3133 Pkg. 1 DCP# lEC-3145 Pkg. 1 "Replacement of Steam Generator Blowdown System Maintenance Isolation Valves (11-14GB3}" -

The purpose of this design change is to replace the existing steam generator blowdown system maintenance isolation valves (11-14GB3} with new valves for better reliability.

Also, the design package modifies five pipe supports inside the containment and four pipe supports in the mechanical penetration area.

(Inside Containment:

lC-BDA-43, lC-BDS-34, 35, lC-BDG-123, 124, Outside Containment:

lP-BDA-95, 97, 98, 99}.

The replacement valve with a higher pressure rating and the modified support designs provide an improved pressure retaining boundary and the improved design of the new valve minimizes the leakage through the GB3 valves during maintenance of the steam generator blowdown system isolation valves (GB4).

This change does not alter the original design intent or the modes of operation or function for which the steam generator blowdown system is currently analyzed.

Therefore, the margin of safety basis for the Technical Specification will not be affected.

(SORC 92-033}

"Installation of Permanent Rigging System for Maintenance of Unit 1" -

The purpose of this design change is to install permanent rigging system (lifting lugs and supplemental steel) for maintenance of the Unit 1 Steam Driven Auxiliary Feed Pump and Turbine, which allows removal and transport of the Pump and Turbine components, at EL. 84' -

O" of Auxiliary Building of Unit 1.

The enclosure will be modified to provide a removable panel.

A bolted panel was chosen, versus a door, to prevent door control problems.

_j

10CFR50.59 EVALUATIONS MONTH: -

MARCH 1992 (Cont'd)

ITEM DCP# lEC-3132 Pkg. 1 DCP# lEC-3112 Pkg. 1 e

DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-272 SALEM 1 APRIL 10, 1992 J. FEST (609)339-2904 The modified opening in the enclosure wall is being assigned a penetration seal number to ensure the seal will not be breached without engineering review and the establishment of proper compensatory actions.

The rigging devices related to the proposed design change and the proposed safe load path do not create the possibility of an accident of a different type than already evaluated in the SAR.

A load drop evaluation was performed and concluded that this drop could not cause any accident that is important to plant safety.

( SORC 92-033)

"Removal of Old and Installation of New NUKON Insulation" -

The purpose of this change is to remove existing metallic reflective insulation and install blanket type NUKON insulation with stainless steel jacketing on the pressurizer vessel, vessel top head and loop seal piping enclosures.

This DCP will also install blanket type NUKON insulation on the bottom head without jacketing.

The loop seal piping enclosures will be removed and replaced with a new structure.

The vessel area covered by loop seal piping enclosure will not be insulated.

UFSAR Section 6.3.2.2 discusses the possibility of insulation material being dislodged and blocking the containment sump screens.

An ECCS Safety Analysis (CD P-501) for Salem per USNRC Regulatory Guide 1.82, revision 1 was performed considering the NUKON insulation being installed by this DCP.

The analysis concludes that the NUKON insulation system will meet the requirements of the Reg. guide when installed in Salem 1 containment.

Therefore, the consequences of a malfunction of equipment important to safety will not be increased.

(SORC 92-031)

"Installation of Monitoring Enhancements" -

The purpose of this change is to install monitoring enhancements to provide early warning of loss of Decay Heat Removal (DHR) capabilities.

This modification will; 1.) Provide at least two independent continuous RCS level indications whenever the RCS is in a reduced inventory condition, and

10CFR50.59 EVALUATIONS MONTH: -

MARCH 1992 (Cont'd)

ITEM DCP# lEC-3099 Pkg. 1 e

DOCKET NO:

UNIT NAME:

50-272 SALEM 1

---

~

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

APRIL 10, 1992 J. FEST (609)339-2904 2.) Provide capability to monitor the residual Heat Removal (RHR) system performance.

The parameters identified for this function are: a.)

RHR Pump Discharge Flow; b.) RHR Pump Discharge Pressure; c.) RHR Pump Suction Pressure; d.) RHR Pump Motor Current, and e.) Mid-Loop Operation core exit thermocouple setpoint.

The changes are designed to safety related criteria and the functional design and safety criteria of interface systems are unaffected.

(SORC 92-031)

"Replacement of 57 Rosemount Electronic Transmitters" -

The purpose of this change is to replace 57 Rosemount electronic transmitters with new transmitters having improved performance under harsh environmental conditions.

The transmitters are: 4 Pressurizer Pressure; 3 Pressurizer Level; 12 Steam Generator; 8 Steam Generator Steam Flow; 12 Steam Generator Pressure; 4 Steam Generator Wide Range Level; 12 Reactor Coolant Flow; 2 Reactor Coolant Pressure.

The new transmitters are seismically and environmentally qualified for the installed locations.

The tubing and valving arrangements will be standardized for all transmitters of the same type.

Where space is available, differential pressure transmitters will be equipped with calibration volume chambers.

These chambers allow calibration with pneumatic signals without the incursion of air into the transmitters or impulse tubing.

The modification also replaces the electronics assemblies on 16 Rosemount transmitters in accumulator level (8) and pressure (8) service.

Each of the 73 transmitters affected by this modification will be equipped with a qualified electrical connector to facilitate maintenance.

There is no change to the instrument process variable spans.

The new transmitters are equal to or better than the previous instruments under all conditions.

Verification of the P-11 allowable value, consistent with the setpoint calculation, will continue to ensure that manual block of safety injection is defeated well above the SI actuation allowable value of ~ 1755 psi.

There is no impact on the margin of safety as defined in the basis for any Technical Specification.

( SORC 92-036)

_j

10CFR50.59 EVALUATIONS MONTH: -

MARCH 1992 (Cont'd)

ITEM.

DCP# lEC-3113 Pkg. 1 DCP# lSC-2267 Pkg. 1 e

DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-272 SALEM 1 APRIL 10, 1992 J. FEST (609)339-2904 "Replacement of Lambda Power Supplies and Daisy Chain" -

The purpose of this change is to replace two obsolete Lambda power supplies with equivalent Lambda power supplies for all five Rod Control Cabinets in Unit 1 and to replace existing daisy chain neutral wiring between the power supplies in a rod control cabinet with individual neutral wires for each power supply.

The rod control circuits which are powered by the new power supplies are not safety related.

The replacement power supplies are equal to or exceed the existing power supplies in all physical and electrical parameters.

The separate neutral wiring from each of these power supplies to the neutral bus will minimize inadvertent tripping of the reactor during changes to any individual power supply when the reactor is at full power.

(SORC 92-036)

"Replacement of Control Electronics Unit in Safeguards Equipment Cabinets" -

The purpose of this change is to replace the existing Control Electronics Unit (CEU) in Safeguards Equipment Cabinets (SEC).

Add a Test Panel to each of the SECs to facilitate monthly functional tests and 18 month timing tests.

Add three undervoltage test switches to each of the SEC cabinets so as to eliminate the need for jumpering during surveillance testing of the 4kV vital buses.

Add a diesel generator start pushbutton (E2) to the existing control panel in the SEC cabinets to facilitate testing.

Revise the procedure to be consistent with new CEU and Test Panel operation including normal operation without ATI engaged.

The reliability of the SEC system is demonstrated by analysis/test results that ensure the modified equipment will withstand the effects of a seismic event.

Reliability is further maintained through the use of MIL-Specification material in the construction of the replacement equipment.

The replacement CEUs interface with existing input and output relays.

No new interfaces are introduced, therefore, the existing redundancy and diversity are maintained.

(SORC 92-036)

10CFR-50. 59 EVALUATIONS MONTH: -

MARCH 1992 (Cont'd)

ITEM DCP# lEC-3143 Pkg. 1 DCP# lEA-1028 Pkg. 1 e

DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-272 SALEM 1 APRIL 10, 1992 J. FEST (609)339-2904 "Permanent Remote Indication of Unit 1 AFP No.

13 Speed and Turbine Inlet Steam Trap" -

The purpose of this change is to provide suitable permanent remote (i.e., external to AFP enclosure indication of Unit 1 AFP No. 13 speed and turbine inlet steam trap drain line temperature as well as AFW discharge piping temperature.

This will eliminate operations personnel entry into a high noise/humidity environment to obtain Surveillance Procedure mandated (as well as INPO recommended) AFP performance parameter measurements.

This proposal will not affect the operability, availability or capacity of 13 AFP or affect the flow path to any Steam Generator as described in Technical Specification 3/4.7.1.2.

Therefore, this modification will not reduce the margin of safety as defined in the basis for any Technical Specification.

( SORC 92-036)

"Reclassify Piping Specification Piping Schedule SPS53D from Nuclear Class III to Non-Nuclear" -

The purpose of this change is to reclassify piping specification Piping Schedule SPS53D from Nuclear Class III to non-Nuclear.

Changes are being made to numerous P&IDs piping class classification tables to incorporate the above changes.

Additional piping class break clarifications and corrections are being made as well.

The other document updates require to incorporate the above changes are UFSAR, S-C-MPOO-MGS 0001-SPS53 and MMIS.

Engineering Evaluation S-C-WD-MEE-0692 documents and accepts the above Nuclear classification changes.

The margin of safety for the Technical Specifications is not reduced since the piping class 53D that is being reclassified is already non-safety related and not part of the Technical Specification bases.

( SORC 92-0 36)

10CFR50.59 EVALUATIONS MONTH: -

MARCH 1992 (Cont'd)

ITEM DOCKET I:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-272 SALEM 1 APRIL 10, 1992 J. FEST (609)339-2904 B.

Procedures and Revisions (Proc)

SC.SA-AP.ZZ-0012(Q)

SP-4 "Technical Specification Surveillance Program" Rev. 15 - This proposal involves a revision to SC.SA-AP.ZZ-0012(Q).

The matrix (Attachment 1 &

2) is being revised to identify that Nuclear Site Maintenance will share the responsibility for performing surveillance number 4.8.3.1 with the Maintenance Department.

Site Services Maintenance (SSM) will be used as a designator for Nuclear Site Maintenance and is being added to the list of department designators.

The revision to the procedure does not change any previously analyzed testing requirement nor does it change any testing method, therefore the proposal cannot reduce the margin of safety as defined in the basis for any Technical Specification.

(SORC 92-031)

"Security Plan Procedure #4" Rev. 2 -

The purpose of this procedure is to revise Security Plan Procedure #4 "The Artificial Island Personnel Access. Program" to describe Security Plan implementation of 10CFR73.56 and assign appropriate responsibility for its implementation based on reorganization of the Site Protection Department to include the Site Access Group.

This revision of the procedure does not reduce the margin of safety because: 1.) It provides a more effective method of administrative ~ontrol over the PAP and other related Security Plan and FFD Program elements, and 2.) It establishes an approach to granting site access which is consistent with other licensees methods of implementing 10CFR73.

Such consistency of approach ensures the temporary workforce is screened to the same standards applied to PSE&G employees, thereby increasing the overall margin of safety.

(SORC 92-036)

SALEM UNIT NO. 1 SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

UNIT 1 MARCH 1992 The Unit began the period operating at full power, and continued to operate at full power throughout the entire period.

REFUELING INFORMATION MONTH: -

MARCH 1992 MONTH MARCH 1992 e

DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

1.

Refueling information has changed from last month:

YES X

NO

2.

Scheduled date for next refueling:

APRIL 4, 1992 50-272 SALEM 1 APRIL 10, 1992 J. FEST (609)339-2904

3.

Scheduled date for restart following refueling:

JUNE 16, 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 656

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