ML18102A983

Monthly Operating Rept for Mar 1997 for Salem Unit 2.W/ 970415 Ltr

ML18102A983
Person / Time
Site:	Salem
Issue date:	03/31/1997
From:	Garchow D, Phillips R Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N970259, NUDOCS 9704220155
Download: ML18102A983 (11)

Text

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• P\Jblic SeNice Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit .~PR 15 1997; LR-N970259 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical specifications, the original monthly operating report for the month of March is being sent to you.

(J;f~/~how David F.

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General Manager -

Salem Operations RAR:tcp Enclosures C Mr. H. J. Miller Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046

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10CFR50.59 EVALUATIONS MONTH: MARCH 1997

• DOC.KE.: 50-311 NAME:

CONTACT:

Unit2 R. Ritzman TELEPHONE: 609-339-1445 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations IOCFR50.59. The Station Operations Review Committee has reviewed and concurs with these evaluations DESIGN CHANGE PACKAGE DCP 2EE-0281, Pkg. 1, Rev. 2 The proposal involves the replacement of a relief valve in the circulating Bearing Safety Evaluation S96- water pump bearing lube water section of the Circulating Water System.

129 Lubrication Pressure Relief The modification is within the Circulating Water System and is not Valve (2SC16) Replacement addressed in the Technical Specifications or "Bases". The effects on the Circulating Water System reactor and its associated margins of safety due to postulated pipe breaks is not addressed in the basis of any Technical Specification section.

Therefore, the proposed modification does not reduce the margin of safety as defined in the basis for any Technical Specifications.

SORC: 97-031 Design Change Package DCP This design change package proposes to install a relief valve upstream of 2EC-3590, Package 1, Revision SW223 (outside containment), with the discharge directed to the common 2, Addition of Thermal service water header downstream of valves SW223 and SW76 for No. 21 Overpressure Device on CFCU and No. 22 header. The SW76 valves will be bypassed for the case when Return Piping the LOCA/MSLB and LOOP transient occurs coincident with a CFCU being out of service. Under this scenario, the applicable CFCU will already be declared inoperable, the SW76 valve will be tagged closed, and the relief valve will discharge to the service water piping downstream of SW76.

SORC: 97-034 Design Change Packages (DCPs) These packages will provide new service water piping and valves within the 2EC-3590, Packages 4&5, Penetration Area for future connection to a proposed accumulator, which Service Water Column will be located outside the Penetration Area. The piping installed under Separation Protection #21 & 22 these packages for the Service Water No. 21 & 22 Nuclear Header supply Nuclear Supply Header piping to the CFCUs. These packages will provide for the main 10" diameter injection piping and the 2" diameter accumulator fill piping (pump suction) within the Penetration Area and terminating just outside the Penetration Area west wall. The piping beyond this point, to the accumulators will be installed in Package 3 of this DCP(2EC-3590) along with the accumulators.

SORC: 97-042

10CFR50.59 EVALUATIONS MONTH: MARCH 1997

• DOCKE.10: 50-311 NAME:

CONTACT:

Unit2 R. Ritzman TELEPHONE: 609-339-1445 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 DESIGN CHANGE PACKAGE Safety Evaluation #97-063 DCP This proposal brings the SW122 valves into conformance with the SW 2EC-3590, Pkg. 12, Rev. 1, System parameters used in the accident analysis. This proposal does not CCHX Inlet Valve SW122 change the SW System acceptance criteria or accident assumptions Volume Booster pertaining to the SW System or any other system. It is, therefore, concluded that this proposal does not reduce the margin of safety as defined in the basis for any Technical Specifications, including Technical Specifications Section 3/4.7.4 (Service Water System), Section 3/4.6.2.3 (Containment Cooling System) or 3/4.3.2 (Engineered Safety Feature Actuation System Instrumentation), as proposed in LCR 96-013.

SORC: 97-048 Safety Evaluation #97-098 2EC-. This modification installs conduit through corebores to the Auxiliary 35.90, Pkg._10,_ Se~i~e \V'~ter:_ _. ~uil,~ing ~a!ls a_n4 routes co_n~µit__~rrgJ.Igh these walls._ The c~mduh is used Accumulator Tanks, Electrical in support of other packages of this design change. The 10CFR50.59 Conduit and Concrete determined that this modification does not increase the probability or Penetrations consequence of any malfunction of equipment or any accident scenarios.

SORC: 97-044 Safety Evaluation #S97-095 DCP This modification addresses concerns regarding the Alternate Shutdown 2EC-3546, Pkg. 2, Rev. 0, Methodology at Salem in the event of a Control Room evacuation due to 10CFR50 Appendix R Rewire of fire in the Control Room, Relay Room, or ceiling area of 460/230 Volt MOV Control Circuits Switchgear Room, which calls for the use of electrical wiring modifications and jumpers. Motor Control Center and control panel circuits for 10 valves will be modified to include local switches which isolate component wiring SORC: 97-048

10CFR50.59 EVALUATIONS MONTH: MARCH 1997

• DOCKE.O: 50-311 NAME:

CONTACT:

Unit2 R. Ritzman TELEPHONE: 609-339-1445 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFRS0.59. The Station Operations Review Committee has reviewed and concurs with these evaluations DESIGN CHANGE PACKAGE Safety Evaluation S97-070 DCP This modification does not reduce the margin of safety as defined in the 2EC-3620 Pkg. 1, Pressurizer basis for Technical Specification 3/4.8.3, Electrical Equipment Protective Heater Circuit Breaker Setpoint Devices, Containment Penetration Conductor Overcurrent Protective Change Devices, rather it increases the safety margin. Technical Specification

3. 8. 3 .1 requires that "All containment penetration conductor overcurrent devices required to provide thermal protection of penetration shall be operable." Improving the coordination of the electrical devices protecting containment penetrations improves the electrical design.

SORC: 97-033 Safety Evaluation S97-072 DCP This modification addresses concerns regarding the Alternate Shutdown 2EC-3546, Pkg. 1, Rev. 0, Methodology at Salem in the event ofa Control Room evacuation due to

'j ciCFRSO Appendix ':R Aiternaie i , fire. itl the Conffolltoom~'Refay Room~ 'or ceiling area of 460/23 o* V oli Shutdown Methodology - Switchgear Room, which calls for the use of electrical wiring modifications Installation of Transfer Switches and jumpers. Motor Control Center and control panel circuits for 10 valves will be modified to include local switches which isolate component wiring from the above stated areas.

SORC: 97-039 Safety Evaluation S97-094 This DCP will modify the existing control actuation circuitry for the Unit 2 Design Change Package (DCP) PORVs. The proposed modification will provide separation/isolation of 2EC-3617, Package 1, Revision safety from non-safety related functions to prevent a single failure from OPORVs Controls Modifications affecting both PORVs.

SORC: 97-046 Safety Evaluation S97-100 DCP The proposed modifications bring the SW system into conformance with 2EC-3590, Pkgs. 3, 6, 11 & 17, the assumptions in -the safety and accident analyses and the *applicable Rev. 0, Generic Letter 96-06 Technical Specifications with regard to the potential for waterhammer.

Modifications The proposed modifications do not change any SW system acceptance criteria or the function of the SW system.

SORC: 97-044

10CFRS0.59 EVALUATIONS MONTH: lVIARCH 1997

• DOCKE.0: 50-311 NAME:

CONTACT:

Unit2 R. Ritzman TELEPHONE: 609-339-1445 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations IOCFRS0.59. The Station Operations Review Committee has reviewed and concurs with these evaluations MISCELLANEOUS Operability Determination 97- This Operability Determination justifies operation in Mode 4 prior to 012 Unit 2 Over-stress Condition correcting the piping configuration of the Unit 2 Steam Generator Water on Steam Generator Water Level Level Instrument Lines. The Operability Determination shows that Instrument Lines operation in Mode 4 is acceptable as long as RCS temperature is maintained less than 240F degrees. Below this temperature, the pipe stress conditions meet design basis requirements.

SORC: 97-035 PROCEDURE Procedure S2.0P-PM.CC- The revision was necessary to permit accurate setting of throttle valves that 0022(Q) - Revision 9, 22 limit the maximum flow through the heat exchangers during accident Component Cooling Heat conditions; however, the revision did not change the actual flow values Exchanger High Flow Flush and during accident conditions. 'The safety evaluation was necessary because Alignment the FSAR stated that the maximum flow through the' CC heat exchangers .

would be limited to a nominal value of 10,000 gpm, but made no mention to of the need have lllgher flows during throttle valve setting/position verification.

SORC: 97-042 Safety Evaluation# S97-078 Performance of this sequencing procedure and all the STPs and plant TS2. SE-SU.ZZ-0001 (Q), Startup manual procedures identified by this sequence procedure will not reduce and Power Ascension Sequencing the margin of safety is defined in the basis for any Technical Specification Procedure because all equipment and systems will be operated IAW the Technical Specifications at all times.

SORC: 97-038

10CFR50.59 EVALUATIONS MONTH: MARCH 1997

• DOCKE"O: 50-311 NAME:

CONTACT:

Unit2 R. Ritzman TELEPHONE: 609-339-1445 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 TEMPORARY MODIFICATION Safety Evaluation S97- This temporary modification 1) isolates the No.21 SW Nuclear Header 027/Temporary Modification #97- 16" supply/return piping to/from the CFCUs to prevent SW flow and 2) if 001Revision2, Installation of required, the provision to isolate the NO. 21 CFCU and No. 22 CFCU 10" Blind Spacers in the No. 21 SW SW supply/return piping to/from containment to maintain containment Nuclear Header CFCU integrity while DCP 2EC-3590 is being worked. The service water supply Supply/Return and the SW to all other users of the No. 21SW Nuclear Supply Header will not be Supply/Return to the No. 21 and impacted once this temporary modification is installed.

No. 22 CFCUs SORC: 97-042 Safety Evaluation S97-028/Temp These spacers may be in place simultaneously in Mode 5 or 6, isolating the Mod 97-002, Rev 2, Installation service water supply to all five CFCUs. However, only one spacer is of Blind Spacers in the No. 22 permitted to'be in place when in MODE 4, providing for the availability of

.SW Nuclear Header CFCU at least two CFC::Us to maintain ~011tainment air temperature when in Mode Supply/Return and the SW 4.

Supply/Return to the No. 23, 24, SORC: 97-042 and 25 CFCUs.

Temporary Modification 97-007 This modification secures open a fire/security door in the Unit 2 Relay Unit 2 Relay Room Pressure Room allowing excess air from the Relay Room to be relieved to the Relief (During Maintenance adjoining stairwell during the maintenance mode of operations. This will Mode of CEACS/EACS lower the pressure in the Relay Room and allow for the required Operation) differential pressure of 0 .125 inwc between the CRE and the Relay Room to be achieved.

SORC: 97-035

• OPERATING DATA REPORT

• Docket No: 50-311 Date: 04/10/97 completed by: Robert Phillips Telephone: 339-2735 Operating Status

1. Unit Name Salem No. 2 Notes

2. Reporting Period March 1997

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

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

11. Hours in Reporting Period 744 2160 136311

12. No. of Hrs. Rx. was Critical 0 0 78083.6

13. Reactor Reserve Shutdown Hrs. 0 0 0

14. Hours Generator On-Line 0 o* 75229.5

15. Unit Reserve Shutdown Hours 0 0 0

16. Gross Thermal Energy Generated (MWH) 0 0 187781005

17. Gross Elec. Energy Generated (MWH) 0 0 78648898

18. Net Elec. Energy Gen. (MWH) -7465 -23197 78625701

19. Unit Service Factor 0 0 54.6

20. Unit Availability Factor 0 0 54.6

21. Unit Capacity Factor (using MDC Net) 0 0 49.0

22. Unit Capacity Factor (using DER Net) 0 0 48.6

23. Unit Forced outage Rate 100 100 32.6

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

Refueling extension.

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

Second quarter of 1997.

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.9:RAGE DAILY UNIT POWER LE~

Docket No.: 50-311 Unit Name: Salem #2 Date: 04/10/97 completed by: Robert Phillips Telephone: 339-2735 Month March 1997 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET} (MWe-NET) 1 0 17 0 2 0 18 0 3 0 19 0 4 0 20 0 5 0 21 0 6 0 22 0 7 0 23 0 8 0 24 0 9 0 25 0 10 0 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14 0 30 0 15 0 31 0 16 0 P. 8.1-7 Rl I

_I

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH March 1997 DOCKET NO.: ~5~0.~*~3~11........._ __

UNIT NAME: Salem #2 DATE: 04-10-97 COMPLETED BY: Robert 'Phillips TELEPHONE: 6~9~339-2735 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 4092 2-1-97 F 744 Fc 4 ...................... 2422 Steam Generator Replacement 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-Refuel ing 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)

Refuel,ing Information Month: March, 1997 Docket No.

Unit Name:

50-311 Salem 2

Contact:

D. Tisdel Telephone: 609-339-1538 Month: March, 1997

1. Refueling information has changed from last month: Yes: X No:

2. Scheduled date for next refueling: Currently in outage.

3. Scheduled date for restart following refueling: To Be Determined

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

Yes: X No: Not Determined to Date:

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

Yes: X (for upcoming cycle) No: If no, when is it scheduled?

4. Scheduled date (s) for submitting proposed licensing action: N/A -

previously submitted

5. Important licensing considerations associated with refueling:

6. Number of Fuel Assemblies:

a. Incore: 193

b. In Spent Fuel Storage: 584

7. Present Licensed spent fuel storage capacity: 1632 Future spent fuel storage capacity: 1632

8. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: October, 2016

pALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 2 MARCH 1997 SALEM UNIT 2 The unit remained shutdown for the entire period. According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, the unit will remain shutdown pending completion of the following actions:

• Appropriately address long standing equipment reliability and operability issues.

• After the work is completed, conduct a restart readiness review to determine for ourselves the ability of the unit to operate in a safe, event free manner.

• After the restart review, meet with the NRC and communicate the results of that review.