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 F, 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 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. RAR:tcp Enclosures C Mr. H. J. Miller (J;f David F.

General Manager -Salem Operations Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046 [\. 0()'J.J -n,v ' .....

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9704220155 970331 PDR ADOCK 05000311 R PDR The power is in your hands. ----------------------.,, 11111111111111111111111111111111 llll llll

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• 95-2168 REV. 6/94 10CFR50.59 EVALUATIONS MONTH: MARCH 1997

• DOC.KE.: 50-311 NAME: Unit2 CONTACT: 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 Bearing Safety Evaluation S96-129 Lubrication Pressure Relief Valve (2SC16) Replacement Circulating Water System Design Change Package DCP 2EC-3590, Package 1, Revision 2, Addition of Thermal Overpressure Device on CFCU Return Piping Design Change Packages (DCPs) 2EC-3590, Packages 4&5, Service Water Column Separation Protection

1. 21 & 22 Nuclear Supply Header The proposal involves the replacement of a relief valve in the circulating water pump bearing lube water section of the Circulating Water System. The modification is within the Circulating Water System and is not addressed in the Technical Specifications or "Bases". The effects on the 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 This design change package proposes to install a relief valve upstream of SW223 (outside containment), with the discharge directed to the common service water header downstream of valves SW223 and SW76 for No. 21 and No. 22 header. The SW76 valves will be bypassed for the case when 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 These packages will provide new service water piping and valves within the Penetration Area for future connection to a proposed accumulator, which will be located outside the Penetration Area. The piping installed under these packages for the Service Water No. 21 & 22 Nuclear Header supply 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: Unit2 CONTACT: 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

1. 97-063 DCP 2EC-3590, Pkg. 12, Rev. 1, CCHX Inlet Valve SW122 Volume Booster Safety Evaluation

1. 97-098 2EC-. 35.90, Pkg._10,_

_. Accumulator Tanks, Electrical Conduit and Concrete Penetrations Safety Evaluation

1. S97-095 DCP 2EC-3546, Pkg. 2, Rev. 0, 1 OCFR50 Appendix R Rewire of MOV Control Circuits This proposal brings the SW122 valves into conformance with the SW System parameters used in the accident analysis.

This proposal does not change the SW System acceptance criteria or accident assumptions 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 This modification installs conduit through corebores to the Auxiliary

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a_n4 routes

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these walls._ The is used in support of other packages of this design change. The 10CFR50.59 determined that this modification does not increase the probability or consequence of any malfunction of equipment or any accident scenarios.

SORC: 97-044 This modification addresses concerns regarding the Alternate Shutdown Methodology at Salem in the event of a Control Room evacuation due to fire in the Control Room, Relay Room, or ceiling area of 460/230 Volt 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: Unit2 CONTACT: R. Ritzman TELEPHONE:

609-339-1445 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 DESIGN CHANGE PACKAGE Safety Evaluation S97-070 DCP 2EC-3620 Pkg. 1, Pressurizer Heater Circuit Breaker Setpoint Change This modification does not reduce the margin of safety as defined in the basis for Technical Specification 3/4.8.3, Electrical Equipment Protective Devices, Containment Penetration Conductor Overcurrent Protective 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

'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 Safety Evaluation S97-094 Design Change Package (DCP) 2EC-3617, Package 1, Revision OPORV s Controls Modifications Safety Evaluation S97-100 DCP 2EC-3590, Pkgs. 3, 6, 11 & 17, Rev. 0, Generic Letter 96-06 Modifications valves will be modified to include local switches which isolate component wiring from the above stated areas. SORC: 97-039 This DCP will modify the existing control actuation circuitry for the Unit 2 PORV s. The proposed modification will provide separation/isolation of safety from non-safety related functions to prevent a single failure from affecting both PORV s. SORC: 97-046 The proposed modifications bring the SW system into conformance with the assumptions in -the safety and accident analyses and the *applicable Technical Specifications with regard to the potential for waterhammer.

The proposed modifications do not change any SW system acceptance criteria or the function of the SW system. SORC: 97-044 lOCFRS0.59 EVALUATIONS MONTH: lVIARCH 1997

• DOCKE.0: 50-311 NAME: Unit2 CONTACT: 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 97-012 Unit 2 Over-stress Condition on Steam Generator Water Level Instrument Lines PROCEDURE Procedure S2.0P-PM.CC-0022(Q) -Revision 9, 22 Component Cooling Heat Exchanger High Flow Flush and Alignment Safety Evaluation#

S97-078 TS2. SE-SU.ZZ-0001 (Q), Startup and Power Ascension Sequencing Procedure This Operability Determination justifies operation in Mode 4 prior to correcting the piping configuration of the Unit 2 Steam Generator Water Level Instrument Lines. The Operability Determination shows that 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 The revision was necessary to permit accurate setting of throttle valves that limit the maximum flow through the heat exchangers during accident conditions; however, the revision did not change the actual flow values during accident conditions.

'The safety evaluation was necessary because 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 of the need to have lllgher flows during throttle valve setting/position verification.

SORC: 97-042 Performance of this sequencing procedure and all the STPs and plant manual procedures identified by this sequence procedure will not reduce the margin of safety is defined in the basis for any Technical Specification 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: Unit2 CONTACT: 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

1. 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 Mod 97-002, Rev 2, Installation of Blind Spacers in the No. 22 .SW Nuclear Header CFCU Supply/Return and the SW Supply/Return to the No. 23, 24, and 25 CFCUs. Temporary Modification 97-007 Unit 2 Relay Room Pressure Relief (During Maintenance Mode of CEACS/EACS Operation)

These spacers may be in place simultaneously in Mode 5 or 6, isolating the service water supply to all five CFCUs. However, only one spacer is permitted to'be in place when in MODE 4, providing for the availability of at least two CFC::Us to maintain air temperature when in Mode 4. SORC: 97-042 This modification secures open a fire/security door in the Unit 2 Relay Room allowing excess air from the Relay Room to be relieved to the adjoining stairwell during the maintenance mode of operations.

This will lower the pressure in the Relay Room and allow for the required differential pressure of 0 .125 inwc between the CRE and the Relay Room to be achieved.

SORC: 97-035

• OPERATING DATA REPORT

• Docket No: completed by: Robert Phillips Date: Telephone:

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) 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 NA This Month Year to Date 744 2160 0 0 0 0 0 o* 0 0 0 0 0 0 -7465 -23197 0 0 0 0 0 0 0 0 100 100 50-311 04/10/97 339-2735 since Last N/A cumulative 136311 78083.6 0 75229.5 0 187781005 78648898 78625701 54.6 54.6 49.0 48.6 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. 8-l-7.R2

.9:RAGE DAILY UNIT POWER Docket No.: 50-311 Unit Name: Salem #2 completed by: Robert Phillips Date: Telephone:

04/10/97 339-2735 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 NO. DATE 4092 2-1-97 F 1 2 F: Forced S: Scheduled DURATION TYPE 1 (HOURS) REASON 2 744 F c Reason A-Equipment Failure (explain)

B-Maintenance or Test C-Refuel ing D-Requlatory Restriction UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH March 1997 METHOD OF SHUTTING LICENSE DOWN EVENT SYSTEM REACTOR REPORT # CODE 4 4 ......................

2422 3 Method: 1-Manual 2-Manual Scram E-Operator Training & License Examination F-Administrative 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther G-Operational Error (Explain)

H-Other (Explain) 4 DOCKET NO.:

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__ _ UNIT NAME: Salem #2 DATE: 04-10-97 COMPLETED BY: Robert 'Phillips TELEPHONE:

--COMPONENT CAUSE AND CORRECTIVE ACTION CODE 5 TO PREVENT RECURRENCE Steam Generator Replacement Exhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 -Same Source * *

• Refuel,ing Information Month: March, 1997

• Docket No. Unit Name: 50-311 Salem 2 Contact: Telephone:

D. Tisdel 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: b. In Spent Fuel Storage: 7. Present Licensed spent fuel storage capacity:

Future spent fuel storage capacity:

193 584 1632 1632 8. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:

October, 2016

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• 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.