ML18096A570

Monthly Operating Rept for Feb 1992 for Salem,Unit 2. W/920312 Ltr

ML18096A570
Person / Time
Site:	Salem
Issue date:	02/29/1992
From:	Shedlock M Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9203200357
Download: ML18096A570 (14)

Text

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• Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station March 12, 1992 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

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 copy of the monthly operating reports for the month of February 1992 are being sent to you. RH:pc cc: Mr. Thomas T. Martin Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions Safety Related Maintenance 10CFR50.59 Evaluations Operating Summary Refueling Information a;;a_ General Manager -Salem Operations Regional Region I 631 Park King of Administrator USNRC Enclosures 8-1-7.R4 Avenue Prussia, PA People ---------

9203200357 920229 PDR ADOCK 05000311 R PDR 19046 95-2189 (10M) 12-89

• 8*-1-7. R2 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-311 Unit Name: Salem #2 Date: 3/10/92 Completed by: Mark Shedlock Telephone:

339-2122 Month *February 1992 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 15 0 31 16 0 P. 8.1-7 Rl

. . P.

• 8. 1-7 Rl *

• OPERATING DATA REPORT Docket No: 50-311 Date: 3/10/92 Completed by: Mark Shedlock Telephone:

339-2122 Operating Status 1. Unit Name Salem No. 2 Notes 2. Reporting Period February 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) since Last Give Reason NA 9. Power Level to Which Restricted, if any (Net MWe) N/A 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) 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 696 0 0 0 0 0 0 -2082 0 0 0 0 0 Year to Date Cumulative 1440 91009 0 58616.1 0 0 0 56898.8 0 0 0 130111721.8 0 59727048 -3901 56864384 0 62.5 0 62.5 0 56.5 0 56.0 100 23.4 24. Shutdowns scheduled over next 6 months (type, date and duration of each) We are presently in a maintenance and refueling outage. 25. If shutdown at end of Report Period, Estimated Date of Startup: April 10,1992 NO. DATE 0001 02/01/92 1 2 F: Forced S: Scheduled DURATION TYPE 1 (HOURS) REASON 2 s 696 Reason A-Equipment Failure (explain)

B-Maintenance or Test C-Refueling D-Requlatory Restriction c UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH FEBRUARY 1992 METHOD OF SHUTTING DOWN REACTOR 4 3 LICENSE EVENT REPORT # -.-----Method: 1-Manual 2-Manual Scram SYSTEM CODE 4 RC E-Operator Training & License Examination F-Aaninistrative 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther G-Operational Error (Explain)

H-Other (Explain)

COMPONENT CODE 6 DOCKET NO. UNIT NAME DATE . COMPLETED BY TELEPHONE 50-311 . Salem #2 03/10/92 Mark Shedlock*

339-2122 . CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE FUELXX NUCLEAR NORMAL REFUELING 4 Exhibit G -Instructions for Prepar.ation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG-0161) 5 Exhibit 1 -Same Source I *

• SAFETY RELATED MAINTENANCE MONTH: -FEBRUARY 1992 DOCKET NO: UNIT NAME: 50-311 SALEM 2 WO NO UNIT 891029014 2 901029033 2 910419262 2 911003013 2 911009202 2 911203120 2 920203154 2 920209085 2 920212081 2 920218147 2 DATE: COMPLETED BY: TELEPHONE:

MARCH 10, 1992 J. FEST (609)339-2904 EQUIPMENT IDENTIFICATION VALVE 2CV181 FAILURE DESCRIPTION:

REPLACE VALVE DIAPHRAGM VALVE 2PR47 FAILURE DESCRIPTION:

REBUILD VALVE ACTUATOR REDUNDANT AIR PANEL FOR 2CV55 FAILURE DESCRIPTION:

REWORK REDUNDANT AIR PANEL FOR 2CV55 VALVE 22SW009 FAILURE DESCRIPTION:

REPLACE VALVE DIAPHRAGM VALVE 2PR1 FAILURE DESCRIPTION:

VALVE 2PR1 LEAKING THROUGH -TROUBLESHOOT VALVE 2CAV16 FAILURE DESCRIPTION:

LIMIT SWITCH FAILED RETEST -INVESTIGATE AND REPAIR VALVE 21CA330 FAILURE DESCRIPTION:

21 CONTROL AIR HEADER NO CLOSED LIMIT -INVESTIGATE 21A & 21B ACCUMULATORS FAILURE DESCRIPTION:

21A & 21B ACCUMULATOR DIFFER -INVESTIGATE

& REPAIR RADIATION MONITOR 2RllA FAILURE DESCRIPTION:

MONITOR 2R11A SPIKES INTERMITTENTLY

-INVESTIGATE SOURCE RANGE CHANNEL 2N31 FAILURE DESCRIPTION:

SOURCE RANGE CHANNEL 2N31 SPIKING -STOPPED FUEL LOAD -INVESTIGATE 10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 *

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

50-311 SALEM 2 MARCH 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# 2SC-2267 Pkg. 1 DCP# 2EC-3125 Pkg. 1 "Safeguards Equipment Cabinet Control Electronics Unit {CEU) Replacement" Rev. 2 -The purpose of this DCP is to replace the existing Control Electronics Unit (CEU) in the Safeguards Equipment Cabinet (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 busses. Add a Diesel Generator start pushbutton to the existing control panel in the SEC cabinets to facilitate testing. This revision turns off the ATI. The reliability of the SEC system is demonstrated by analysis/test results that ensure the modified equipment will withstand the affects 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.

Therefore, this DCP does not reduce the margin of safety as defined in the bases of the Technical Specifications.

{ SORC 92-012) "Lube Oil Flushing Modifications" -This change involves the installation of a flushing connection to the main turbine oil reservoir.

The flushing connection will only be used with the plant in the shutdown mode. Since this change is for the installation of a connection made to the reservoir to facilitate flushing which will be accomplished with the plant in the shutdown mode, it will have no affect on accidents or malfunctions during plant operation.

1--.-1 I I I I I I I I I I I I I I I I I 10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 (Cont'd) ITEM

• DCP# 2EC-3106 Pkg. 1 DCP# 2EC-3087 Pkg. 1

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 It also will have no affect on accidents or malfunctions analyzed for the shutdown modes. Therefore, there is no reduction in the margin of safety as defined in the bases for any Technical Specifications. (SORC 92-012) "Replacement of Damaged Condenser Tubes" -This change requires the replacement of damaged condenser tubes included in Sections 21B, 22A, and 22B, of the Main Surface Condenser.

The tube material is being upgraded from AL6X to AL6XN. This change also includes the replacement of a portion of the 2-inch bleed steam drain line in sections 21B and 22B of the Condenser, due to erosion damage and replacement of the 3-inch MSR main steam coil drain tank drain line spray headers in Condenser No. 22. One header was damaged during the turbine incident, and the other is damaged due to erosion. The material is being upgraded from carbon steel and 1-1/4 chrome-1/2 molybdenum, respectively, to 2-1/4 chrome-1 molybdenum.

Repairs to the condenser and auxiliaries also includes demolition and repairs to structural supports, bleed steam and miscellaneous piping, and the condenser steam inlet expansion joint. These repairs are being accomplished via engineering work packages and are outside the scope of this DCP. This DCP involves the replacement of equipment damaged during a turbine failure incident.

The function of the equipment is unchanged.

The equipment being replaced is non-safety related and serves no safety related function. ( SORC 92-014) "RHR Monitoring During Mid-Loop Operations" -This change package provides for installation of monitoring enhancements to provide early warning of Loss of Decay Heat Removal (DHR) capabilities.

This instrumentation provides a monitoring, indication and alarm function only and performs no active control of any plant equipment.

10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 (Cont'd) ITEM

• DCP# 2EA-1012 Pkg. 1 DCP# 2EC-3124 Pkg. 8 DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 The circuitry up to and including the isolation device is safety related and qualified in accordance with IEEE 323 and IEEE 344 as appropriate. ( SORC 92-014) "23 Vacuum Pump Blind Flange" -The purpose of this change is to permanently document the installation of a blind flange isolating 23 Vacuum Pump from the air removal header and designating it as the "prime only" pump. The proposed change is non-nuclear and non-safety related and has no affect on the operation of the system. The condenser and air removal system is a secondary system. Documentation of the installation of the blind flange does not increase the probability or consequences of an accident previously evaluated in the SAR. The plant can reject heat without the use of the ,condenser and has been previously evaluated in the SAR within the discussion of a loss of offsi te power. ( SORC 92-022) "Rerouting of Stator Cooling System Piping" This DCP involves the rerouting of existing stator cooling system piping and generator lead box drain pipes to match the connections to/from the new replacement generator.

This modification does not change system design, function, or operation.

Additionally, this system is not required for the safe shutdown of the plant. Stator cooling water flow requirements for the replacement generator will not be increased, according to General Electric.

The additional pressure drop due to the piping modification is approximately 1.35 psi. This is acceptable because the cooling pump has an additional margin to accommodate this increased pressure drop without affecting the required flow rate. (SORC 92-018) 10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 {Cont'd) ITEM

• B. Procedures and Revisions VS2.RE-FR.ZZ-0003(Q)

TS2.0P-SO.ZZ-0008(Q)

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 "Westinghouse Refueling Procedure

-FP-PNJ-R6" Rev. 3 -This procedure is revised to provide steps in the Insert Changeout Section and the steps necessary to reload the core. These steps are being added due to the reevaluation of the core design following the Ultrasonic/Visual Inspection performed on the Irradiated Fuel. Also, a statement has been added to the precautions and limitations which allows the bridge formed across the core to be straight or semicircular.

This procedure will be performed in accordance with existing site procedures, and the requirements of Technical Specification 3/4.9, Refueling Operations, will be maintained at all times. By following approved existing site procedures and processes, and maintaining the requirements of the Technical Specifications, the margin of safety.for a full core reload will be bounded by the existing analysis for a core shuffle as described in UFSAR Section 9. ( SORC 92-014) "Remove/Return From/To Service the 2A-125 VDC Bus" -This new procedure will (1) provide instructions to remove/return from/to service the 2A-125 VDC Bus during Modes 5 and 6, (2) ensure that all required loads are energized from their backup DC power source, (3) ensure minimum system disruption during removal/ returning, and (4) ensure compliance with Modes 5 and 6 Technical Specifications

[i.e., Two{2) out of three(3) Vital 125 VDC Busses are operable.

Redistribution of the 2A-125VDC battery loads will (1) still maintain adequate DC power to connected loads (2) not increase battery sizing requirements, based on the System Study (3) not cause a LOPA previously evaluated in the SAR (4) still allow the battery to operate as designed, and (5) still allow all safety related systems that use this 125VDC power , to operate and maintain their original margin of safety as defined in the Technical Specification Bases. ( SORC 92-014) 10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 (Cont'd) ITEM

• VS2.MD-PM.SW-0001(Q)

DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 "Application of Ceramalloy Cladding" -Pump Room coolers 21RHR, 22RHR, 21CC, 22CC, 2Sl & 21AFW will have their tube sheets coated with a ceramic polymer coating to stop further crevice attack and enhance leak tightness.

The coating to be used is compatible with system materials.

The engineering calculations do not take credit for the heat transfer performance of the tube sheet. Thus if the coating affects its capability it is of no significance.

The cladding is considered to be a crevice corrosion inhibitor and leak tightness augmenter for the tube/tube sheet interface.

The proposal does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 92"-024) C. Temporary Modifications (TMOD) TMR 92-015 D. Safety Evaluations (S/E) NFU 92-072 Removing/Returning 2A 125VDC Bus From/To Service" -The purpose of this TMOD is to provide temporary power to 23 Charging Pump, during the 2A Bus Outage to make it Operable as per the Technical Specifications.

This modification installs temporary jumper cables between 480V AC Vital Busses "2A" & "2C". Any failure of this jumper cable could result in a loss of both these busses. However, it must be noted that during the time this TMOD is installed, Vital Bus "2A" is already inoperable.

Therefore, any failure of this jumper cable will only make one Vital Bus "2C" inoperable.

The same failure mode exists for the present lineup and no new failure modes are created due to this TMOD. (SORC 92-015) "Salem Unit 2 Cycle 7 Core Loading Pattern and Safety Evaluation for Operation in Modes 6, 5 and 4 -The safety evaluation for the reload and operation of Salem Unit 2 cycle 7 has been completed.

During the refueling outage issues have been identified regarding Auxiliary Feedwater flow rates, AFW turbine pump cavitation, and Containment Spray Delay time.

EVALUATIONS MONTH: -FEBRUARY 1992 (Cont'd) ITEM S-2-RC-NSE-0805

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 The complete RSE for Salem Unit 2 cycle 7, covering all modes of operation will only be applicable following resolution of the AFW and Containment Spray issues. This RSE is applicable in Modes 6, 5, and 4. As a result of the reload analysis, no new failure modes or limiting single failures have been identified for Salem Unit 2 Cycle 7 reload core design operation in Modes 6, 5 and 4. The increased containment spray delay time and the inoperability of the Auxiliary Feedwater System in Modes 6, 5 and 4 will not adversely affect the performance of any other equipment and does not change the design basis for any other equipment.

Since the design basis for the containment spray system is not changed and the AFW system is not required to be operable in these modes, there can be no initiation of any accident or creation of any new credible limiting single failure. ( SORC 92-014) Foreign Materials in the Reactor Vessel" -A piece of lucite with size approximately l" x 1/2" x 1/4" was lost in the Salem Unit 2 Reactor Vessel during a recent in-service inspection activity.

A subsequent foreign object search and retrieval effort was not able to locate it. The fuel assemblies and the lower internals were not in the vessel when the object was dropped. Therefore, after the lower internals and and fuel are placed into the vessel, the object will remain below the active core region. Its size will not allow it to migrate into the fuel assemblies.

Studies using the THINC-IV code have shown that substantial blockage (10%) of a fuel assembly inlet does not result in significant reductions in assembly flow due to cross flow recovery.

Since any potential inlet flow blockage due the missing object would only be approximately 1% of the assembly flow area, there will be no effect on LOCA calculated peak clad temperature.

In addition, the piece of acrylic is expected to dissolve during plant operation due to high temperatures and radiolytic deterioration.

--.---.---

10CFR50.59 EVALUATIONS MONTH: -FEBRUARY 1992 (Cont'd) ITEM *

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

SUMMARY

50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904 Therefore, the piece of acrylic will not adversely affect the Salem UFSAR Chapter 15 safety analyses.

The object does not create a corrosion concern for RCS materials nor does it create a contamination concern for primary side water chemistry. (SORC 92-017)

• SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

-UNIT 2 FEBRUARY 1992 SALEM UNIT NO. 2 The Unit was out of service for the entire period for the Sixth Refueling Outage.

REFUELING INFORMATION MONTH: -FEBRUARY 1992 MONTH FEBRUARY 1992 *

• DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:

1. Refueling information has changed from last month: YES X NO 50-311 SALEM 2 MARCH 10, 1992 J. FEST (609)339-2904

2. Scheduled date for next refueling:

NOVEMBER 11, 1991 3. Scheduled date for restart following refueling:

APRIL 15, 1992 4. a) Will Technical Specification changes or other license amendments be required?:

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

YES x NO 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 408 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:

March 2003 8-1-7.R4 * -Refueling outage dates may be revised due to turbine generator failure.