ML20236Q708

Monthly Operating Rept for Oct 1987

ML20236Q708
Person / Time
Site:	Hope Creek
Issue date:	10/31/1987
From:	Jensen H, Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8711200239
Download: ML20236Q708 (22)

Text

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l AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.86-354 UNIT Hope Creek DATE 11/13/87

> COMPLETED BY H.'Jensen TELEPHONE (609) 339-5261 MONTH October 1987 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POfr LEVEL !

(MWe-Net) (MWe-Net) l' 0 17 0 2 0 18 0 3 0 19 0 I

4 0 20 0 5 0 21 0 6- 0 22 0 7 'O 23 0 8 0 24 0

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9 0 25 0 10 0 26 0  !

11 0 27 52 l

12 ,

0 28 751 13 0 29 945 i 14 0 30 1048 I 15 0 31 1021 16 0 DR 39 871031 /

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DOCKET NO.86-354 1 i

7 UNIT Hope Creek j DATE 11/13/87 COMPLETED BY. H. Jensen TELEPHONE' (609) 339-5261 MONTH. October 1987

. DAY AVERAGE' DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

'l 0 17 0

'2 0 18 0 ,

.3 0 19 0 4 0' 20 0

5. 0 21 0-6 0 22 0 l 7 0 23 0 8 0 24 0 9 0 25 0 10 0 26 0 11- 0 27 52 12 0 28 751 13 0 29 945 14 0 30 1048

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OPERATING DATA REPORT DOCKET NO.86-354 UNIT Hope Creek DATE 11/13/87 COMPLETED BY H. Jensen Ah -

TELEPHONE (609) 339-5261 OPERATING STATUS

1. REPORTING PERIOD Oct 1987 GROSS HOURS IN REPORTING PERIOD 745

2. CURRENTLY AUTHORIZED POWER LEVEL (MWt) 3293 MAX. DEPEND. CAPACITY (MWe-Net) 1067

• DESIGN ELECTRICAL RATING (MWe-Net) 1067

3. POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net) None

4. REASONS FOR RESTRICTION (IF ANY)

THIS YR TO MONTH DATE CUMULATIVE

5. NO. OF HOURS REACTOR WAS CRITICAL 175.1 6136.2 6424.2

6. REACTOR RESERVE SHUTDOWN HOURS 0 0 0

7. HOURS GENERATOR ON LINE 107.0 6027.7 6315.7

8. UNIT RESERVE SHUTDOWN HOURS 0 0 0

9. GROSS THERMAL ENERGY GENERATED (MWH) 310,174 18,221,808 19,152,216

10. GROSS ELECTRICAL ENERGY GENERATED (MWH) 101,422 6,059,015 6,356,655

11. NET ELECTRICAL ENERGY GENERATED (MWH) 86,230 5,786,859 6,072,683

12. REACTOR SERVICE FACTOR 23.5 84.1 84.7

13. REACTOR AVAILABILITY FACTOR 23.5 84.1 84.7

14. UNIT SERVICE FACTOR 14.4 82.6 83.3

15. UNIT AVAILABILITY FACTOR 14.4 82.6 83.3

16. UNIT CAPACITY FACTOR '

(Using Design MDC) 10.8 74.3 75.0

17. UNIT CAPACITY FACTOR (Using Design MWe) 10.8 74.3 75.0 l

18. UNIT FORCED OUTAGE RATE 79.0 10.7 10.3

19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, & DURATION):

Refueling 2/12/88, 55 days

20. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:

• Data obtained in August is under management review. l

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OPERATING DATA REPORT UNIT SHUTDOWNS-AND POWER REDUCTIONS DOCKET NO.86-354 UNIT Hope Creek DATE 11/13/87 COMPLETED BY R. Ritzman REDORT MONTH Oct. 1987 TELEPHONE (609) 339-3737 METHOD OF SHUTTING DOWN THE.

TYPE . .

REACTOR OR

.F FORCED DURATION REASON REDUCING CORRECTIVE ACTION /

'HO. DATE. S SCHEDULED. (HOURS) (1) POWER (2) . COMMENTS 17 10/1 S 235.8 B~ 4 Continuation of the Planned Surveillance Outage 18 10/10 F 60.4 A 2 "J" Safety / Relief Valve failed to close LER 87-047 10/10/87 19 10/13 F 339.0 A 1 Main Transformer Fire 20 10/27 F 2.8 A 1 Turbine and  ;

Generator offline l b for electro-hydraulic control system maintenance

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HOPE CREEK ~dENERATING STATION MONTHLY OPERATING ~

SUMMARY

OCTOBER 1987 The planned' surveillance outage continued through the beginning of the month until October 10'when fplant 'startup was. commenced. During startup' testing of the safety / relief valves- to obtain baseline data for the. acoustic' monitoring system, the "J" safety / relief valve failed to close. 'The reactor was manually scrammed at 7:49'PM on October 10.

The reactor again went' critical at 7:56 AM on October 12 and was manuallyfsbut down due to a fire in.the "A" Main' Transformer following

'/'~ synchronization'with'- the. grid for approximately one. hour. After replacing'the "A"' Main Transformer, the unit went critical at 9:16 AM on October 26; synchronizing the Main Generator with the grid at.11:35 AM:on the 27th. The turbine and generator were taken off-line at 6:22 PM for Electro-Hydraulic Control system maintenance, with the . reactor remaining critical. The Main Generator was again synchronized with j the' grid at 9:11 PM' on October 27. At months end the' generator 'had-  :

been on-line for 4 days.

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SUMMARY

OF. CHANGES, TESTS,-AND EXPERIMENTS I FOR THE HOPE CREEK GENERATING STATInN OCTOBER 1987 t

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DCP Description of Desian Chance Packace 1

.4-ECM-86-0226B ~ This DCP installed- blowout panels, in the Containment Prepurge Cleanup system ductwork Land revised setpoints for the torus compartment .and l ' interconnected rooms isolation ' damper switches.

This DCP supports the ' qualification of' the

' operability of the Containment Atmosphere Control system during plant operation.

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The increase- of the pressure switch setpoints  !

' delay the. closing of the isolation dampers-in the I event of' a. pipe breax, which increases the spreading of Loss of Coolant Accident fluid into the. Filtration, Recirculation, and Ventilation system and the Reactor Building Ventilation system ductwork. These isolation dampers may also be closed independently 'due to the closing of temperature switches local to the pipe break. The increase in the consequences of. an accident -are'

-small and acceptable.

The operation of the . vent / purge containment-isolation valves during plant operation . increases the probability of exposing the Filtration, Recirculation,- and Ventilation system ductwork and equipment to increased pressure as well as containment atmosphere during a Loss of Coolant Accident. This DCP has been designed to limit.the affect of such an accident.

By permitting the release of Loss of Coolant fluid.

into the connecting rooms through the blowout panels, the margin of safety may be decreased.

All consequences of such an accident have been evaluated to ensure that they have a negligible affect on plant safety.

The NRC has previously reviewed these changes and ,

concurs that the required level of safety is maintained.

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The following Desi'gn Change Packages (DCPs) have been evaluated to determine:

1) if the -probability of occurrence or .the- consequences of- an-accident or malfunction 1 of equipment- important to' safety previously_ evaluated in the. safety' analysis report may be

. increased; or a

2). if-a possibility for an accident or ' malfunction of a different-type than any evaluated previously in the safety analysis report may be created;.or

'3) iif the' margin of safety as defined in the basis for any technical specification is reduced.

~None of the DCPs created a'new safety haza'rd.to the plant'nor did they affect:the safe shutdown of the reactor. These DCPs did not change-the plant. effluent releases and did not alter the existing environmental impact. The Safety Evaluations determined that. no'  ;

unreviewed safety or environmental. questions are involved.

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TDCP 1

681 This DCP'replacedL all non-Rockbestos cable. going.

to the Control Room' Integrated Display System and.

Nuclear Steam Supply System computer keyboards and

-terminals with Rockbestos cable. This DCP did not.

change the system logic, the circuitry, or operation of'the~ equipment. All ofthe.' material' that was used is qualified for its application and was installed in order'to meet-plant requirements.

.HC-QF-002 This DCP.' added speakers- and- handsets for the Public' Address system. They were added to provide improved in-plant. communications to- Heating and Ventilation Equipment Rooms, elevation 102' Reactor Building, Radwaste Control Room, System Engineering office area, Tagging Office Control Room Complex Group office area, and the Switchyard i Control Building. . This. DCP also added a

" priority" page'to allow the Control Room to make

., uninterrupted announcements. The wiring changes were1 evaluated to and comply with 10CFR50 Appendix R, electrical separation,' and seismic II/I i criteria.

4-FMJ-86-0941' .This DCP installed modens and cards in computer cabinets. It also added computer points and made nomenclature changes so the Radiation Monitoring System database matches the Control Room Integrated Display System database. This DCP enhanced the information available to station personnel. No safety-related equipment or control l functions were affected.

HMM-86-lll6 This DCP replaced 3/8" copper tubing with bronze / brass flexible hose in Instrument Air Supply Lines in the Fuel Pool Filter Demineralized System. The change was necessitated by vibration during operation and enhances the operation of the associated valves.

4-HCE-86-1214 This DCP changed the contact status of scram contacts from normally closed to normally open.

This change assures reliability of the Control Room Integrated Display System and annunciation on either a scram or loss of power to the digital isolators. This DCP affects annunciation only and restores the original design intent.

4'-HME-86-1251 This DCP revised the time delay setpoint to the Generator Field Ground Detector Relay from 1 second to 1/2 second. This change provides a faster alarm and trip in the event of a generator

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field ground which, in turn, lessens the possibility and consequences of equipment damage.

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DCP: Description'of Desian Chance Packace 4-HMM-86-1260 The Condensate Demineralizers. must regenerate resins after only 50% capacity is used. However,

.theiRadwaste Regeneration. System can utilize. the .

' resins to' .a significantly higher- capacity. l Therefore, this DCP added a valve and a'Eremote gear. operator-in order to isolate the Spent Resin Tank' during the transfer. of resins from the Condensate Demineralizers to the- Radwaste Regeneration System, allowing greater .use to be made of the resins.

4-HMJ-86-1269 'This DCP added a Reactor Auxiliaries Cooling System Water ' Permissive Signal to the' Emergency.

Air' Compressor Auto Start Circuitry in :the Instrument Air system. This change . prevents unnecessary compressor' ' starts, increasing system-

l. reliability.

4-EC-1055/l' 'This DCP . removed equipment and non-structural steel-from the rotating Control Rod Drive Handling Platform. The removal of this equipment and steel-is necessary to allow the later-installation of a semi-automatic Control Rod Drive Handling Machine, which will significantly reduce personnel.

radiation exposure.

4-EC-1055/2' This DCP installed new rails and support structure components in preparation for the installation of a -semi-automatic Control Rod Drive . Handling i Machine. All changes made by this DCP have been- '

stress analyzed and comply with the original.

design criteria for this component.

4-HC-0056 This DCP provided an alternative method of I isolating the Auxiliary Steam System from- the Reactor Feed Pump. Turbines by installing blind i flanges in the Auxiliary Steam Lines. This change does not alter the primary method of isolating the Auxiliary Steam Lines, it merely provides an acceptable alternate.

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Descrit> tion of Desian Chance Package; b .

4-HC-0143/1 This DCP 'provided for -the deletion: of the stem  !

drain and bonnet vent connections to. the B Loop )

Recirculation Suction and Discharge valves. .These '

modifications.will eliminate the potential for ~

piping' failure due to vibration and the subsequent steam leakage. A review of valve function as a-1 result. of these deletions shows 'that valve function and performance remain unchanged.

1 4-HC- 314 3/2 This DCP .provided for the deletion of the ' stem' H drain and bonnet vent. connections to the AL Loop j Recirculation Suction and Discharge valves. These modifications will eliminate the potential'.for piping failure due to vibration and the subsequent system leakage. A review of valve function as a result of these deletions shows that' valve function and, performance remain unchanged.

4-HC-0143/3 This DCP .provided for the deletion of the' low point drains of the reactor ' recirculation loop flow differential pressure sensing instrumentation lines connected to the outer. elbow taps of the B j loop recirculation suction elbow. These I modifications will aid in minimizing the potential for- piping failure due to vibration and the subsequent system leakage. Deletion'of the valves n will not alter the function of the instrument lines because alternate methods of line flushing are available. The amount of undrainable fluids present'when the loop would be completely . drained isi minimal. This- amount can be virtually eliminated by blowing air'through the drywell vent of the sensing line thus pushing the undrainable water into the recirculation elbow and through the recirculation suction valve seat drain.

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DCP Description of Desian Chance Packace 4-HC-0143/4 This DCP provided for the deletion of the low point drains of the reactor recirculation loop flow differential pressure sensing instrumentation lines connected to the outer elbow taps of the A loop recirculation suction elbow. These modifications ill aid in minimizing vibration and the subsequent system leakage. . Deletion of the valves will not alter the function of the instrument lines because alternate methods of line flushing are available. The amount of undrainable 5 flui6 present when the loop would be completely drained is minimal. This amount can be virtually eliminated by blowing air through the drywell vent of the sensing line thus pushing the undrainable water into the recirculation and through the recirculation suction valve seat drain.

4-HM-0001 This DCP installed pressure gauges and assemblies on Turbine Control Oil Hydraulic Accumulators for the Reactor Feed Pump Turbines. This change increases the gauge efficiency and decreases the time required to che<k and/or adjust accumulator pressure.

4-HM-0010 This DCP' added isolation valves and test "T" fittings in line with the Reactor Protection System Control Valve Closure pressure switches.

The addition of the valves and fittings facilitate routine calibration and maintenance activities.

4-HM-0043 This DCP changed the internal parts of the Circulating Water Caustic Injection Pumps from Viton to Hypalon. The system was changed from an acid injection system to a caustic injection system; however, the internals were never changed-out. Hypalon is more compatible with caustic solutions than Viton is, therefore, this change will reduce the failure rate of the pump internals.

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l 4-HM-0086-

'This_DCP added a capacitor to the backplane of a-l display memory -module in the Reactor Manual Control System. This change: provides filtering h ' for the synchronization line ,to the Rod Data '

Memory Cards and will ' eliminate 'the. random flashing of. rod status-on the_. full core display.

i 4-HM-0112 This DCP. replaced the Bisulfite Line in- the

c Cooling- Tower . Blowdown Dechlorination system i becausenof a blockage in the current line. It also added vent / flush point ' connections at each .

- end of the line. The. vent / flush points will allow I future blockages to be cleared.

4-HM-015 5 This-DCP changes the level sensor relay for- the s Radwaste Boiler Low Level Cutoff Switch to a more  :

sensitive . relay. This. change will allow the Radwaste Boiler to operate more reliably hnd with less operator attention. The new switch is better' suited for demineralized water, which is used in this application.

4-HM-0173 This DCP added orifice plugs to the Main Turbine Electro-Hydraulic Control Emergency Trip System pressure ports at each Fast Acting Solenoid Valve.

The orifice ~ plugs will prevent pressure transients during routine Turbine Control Valve

' surveillance.

4-HM-0174 This DCP utilized spare contacts on the Electric-Overspeed Trip Relays as an enable for the Turning j Gear- Zero Speed Switch. This change will l eliminate sporadic Control Room ' alarm annunciations due to intermittent activation of the Turning Gear Permissive logic.

4-HM-0187- This DCP changed the low power setpoint and alarm point on the Rod Sequence Control System.  !'

Changing the setpoints incorporate an additional margin to account for instrument accuracy and uncertainties in Turbine First Stage Pressure under various plant conditions. This change restores the system to the design condition.

4-HM-0192 This DCP modified the joint design of the Cooling  ;

Tower Distribution Flumes. The original design j utilized expansion joints which resulted in a generic failure. The new design utilizes a rubber seal which will prevent leakage while allowing for thermal expansion.

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'DCP- Description of Desion Chance Packace.

-42HM-0200 ThisfDCP ? improved the Reactor. Protection System grounding scheme.by removing internal panel wires.

.These' wires 1 were originally installed tx) -ground regulating isolation transformers. . Subsequently, the transformers were removed but the wires were noti . Removing 'them. reduces the likelihood. of ground loop. problems..and improves the. system reliability.

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'The' following- Temporary- Modification Requests (THRs). have been t aluated to determine:

1) if the. probability .of occurrence'or the consequences oof .an accident or malfunction of equipment important to safety previously evaluated in .the safety analysis report may be increased; or

2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or 3)' if the margin of safety of safety as defined in.the basis for any

. technical specification is reduced.

None:of the TMRs created a new safety hazard to.the plant nor did they affect the safe shutdown of the reactor. These TMRs did not change the plant effluent releases and did not alter the- existing environmental impact. The Safety Evaluations determined that no unreviewed safety or environmental questions are involved.

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P Safety Evaluation Description of Temporary Modification Recuest (TMR)

H-1-BBXX-MSE-0710 This~ TMR removed equipment and added pipe vibration instrumentation to the Reactor Recirculation' system. This instrumentation will monitor steady state vibration on the suction lines and associated instrument lines where cracks were discovered.

. H-1-MXXX-CSE-0713 This TMR installed temporary instrumentation to monitor and record the Main Generator and Main Transformer volts / amps / phase angle to determine ,

megawatt output discrepancies between the Watt  !

transducer and the watt meter circuits. All circuitry covered by this TMR is contained within the generator / transformer relaying cabinets and does not interface with safety-related systems or components.

87-0139 This TMR removed overload heaters from the breaker for a valve in the Residual Heat Removal system.

This will prevent the valve from opening during a 10CFR50 Appendix R fire.

87-0140 This TMR lifted wires to several solenoids in the Residual Heat Removal system. 10CFR50 Appendix R requires that power be removed to prevent inadvertent operation of the valves during a fire.

87-0141 This TMR lifted leads in several panels to remove the power source from the Core Spray testable check valve and bypass valve solenoids. Removing the power complies with 10CFR50 Appendix R.

87-0173 This THR removed overload heaters from the breaker for a valve in the Residual Heat Removal system.

.This will prevent the valve from opening during a 10CFR50 Appendix R fire.

87-0176 This TMR removed overload heaters from the breaker for a valve in the Residual Heat Removal system.

This will prevent the valve from opening during a 10CFR50 Appendix R fire.

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.t Safety Evaluation Description of-Temporary Modification.

Reauest (TMR) 87-0179. This TMR installed Teflon thermocouple wiring instead of Kapton wiring on the "B" Reactor Feed Pump. The-' existing Kapton wiring failed and no l replacement is'immediately available. The Teflon i wiring does not preclude the Reactor Feed Pump i from fulfilling its designed function.

.87-0180 This TMR replaced the Fuller Earth Filter Inlet and Outlet Pressure Indicators with an equivalent ..

gauge. This -has no functional effect on' the i Electro-Hydraulic Control System.

87-0181- 'This.TMR replaced a Westinghouse logic card in the Service Water Pump Panel with one of a different i model. Both cards- were tested to the- same criteria and are. compatible in this application.

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4 The following Deficiency Report (DRs) have been evaluated to determine:

1) if the probability of occurrence or the consequences of an

-accident'- or malfunction of equipment important to safety.

previously evaluated in the safety analysis report may be

-increased; or

2) if a possibility for an accident or malfunction of a different

' type than any evaluated previously in the safety analysis report may be created; or

3) if the margin of safety of safety as defined in the basis for any technical specification is reduced.

None of the DRs created.a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These DRs did not change the

-plant' effluent releases and did not alter the. existing environmental impact. The Safety Evaluations determined that no unreviewed. safety or environmental 1 questions are involved. {

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-Safety Evaluation Description of Deficiency Report (DR) 87-011-3' This Safety Evaluation addresses several problems associated with- the Radiation Monitoring ' System Computer. The computer has no control functions and none of these problems impaired its ability to provide- the required alarms, statuses, .and indications to the Radiation Monitoring System 3 Monitors. ]

87-0114 The Radiation Monitoring System computer did- not properly indicate the Suppression Pool Temperature. The Suppression Pool Temperature e Monitoring System monitors are operating properly and meet-the Technical Specification requirements for monitoring, alarming, and recording.

'87-0116 The Drywell Leak Detection - Cooler condensate Monitor could.not have its database automatically reloaded from the Radiation Monitoring System.

Alternate. methods exist for reloading the data.

These methods .are. currently being utilized .and therefore continue to meet the design requirements.

87-0120- The Hydrogen Oxygen Analyzer Trouble Light in the Main Control Room is in constant alarm. This is

. caused by the flow switches and the sealed volume switches having operating ranges outside of the unit's operating pressure. This problem does not affect operability because both flow and sealed volume can be verified at a local panel. A DCP will be implemented to alleviate this problem.

87-0124 The temperature bulb for the Chilled Water Low Temperature Cutout Switch is installed incorrectly. The chiller unit also has an Evaporator Low Refrigerant Pressure Cutout Switch which provides a redundant function. Due to this switch, the ability of the chiller to fulfill its designed function is not compromised.

87-0126 The Turbine Building Compartment Exhaust and Turbine Building Exhaust Ventilation Duct Radiation Monitors are in constant alarm due to radiation shine from the Main Turbine. These monitors are used only to alarm and to provide trending. They have no isolation functions.

Additionally, grab samples may be obtained to determine the amount of airborne radioactivity that exists in the ductwork.

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Safety Evaluation Description'of Deficiency Report (DR) 87-0127 The upper travel . stops for the Fuel- Preparation Machines in the-Spent Fuel Pool are set too high.

.These stops -are required. to ensure proper radiation shielding for irradiated fuel. There is not irradiated fuel in the spent Fuel Pool at this time and the stops will be set correctly prior- to  ;

the handling of irradiated fuel.

!87-0142 The motor-operated. valve providing flow- isolation of the Safety Auxiliaries. Cooling System-to the Residual Heat Removal Heat -Exchanger leaks through. This valve is open during a Loss of Coolant Accidant and. ' closed during normal operations. Diverting this flow (approximately 5%

of capacity) from the Turbine Auxiliaries Cooling System to the Residual. Heat Removal Heat Exchanger during normal operations does.not adversely affect either system.

87-0147 The cable to the Safety Auxiliaries Cooling System Water Supply Valve to the- Filtration, Recirculation, and Ventilation System Cooling Coil-unit is damaged. This cable will be' repaired so as to- restore the~ cable to 'its original design specification.

87-0148 The vibration level on a Standby Liquid Control Pump exceeds the procedural limits. The vibration level is satisfactory; however,'the procedure did not identify the optimum points for vibration monitoring. The . procedures will be revised to reflect the correct method of vibration monitoring.

87-0149 A p2 essure relief valve in the Hydrogen Oxygen Analyzer leaks through. The leakage occurs when the calibration gas is valved in. Since the calibration gas is required only during calibration, it will be valved out during normal use. This both eliminates the leakage problem and assures the availability of the calibration gas when it is required.

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-Safety-Evaluation' Description of Deficiency Report (DR) 87-0151 Check valves are used to supply water pressure from the Condensate Storage and Transfer system to the discharge of the High Pressure Coolant

' Injection and Reactor Core Isolation Cooling piping in the event of.their jockey pump failure.

Due to slight packing friction, these valves might not fully seat. However, these check valves are normally isolated'from the High Pressure Coolant Injection and Reactor Core Isolation Cooling systems by a. globe: valve which provides a .

redundant function in'this application.

i 87-0152 Check valves are used to supply water pressure .;

from the Condensate Storage and Transfer system to.

the discharge of the Core Spray and Residual -Heat  !

Removal piping.in'the. event of their' jockey pump failure. Due to slight packing friction, these valves might not fully seat. However, these check q valves are normally isolated from the Core Spray l and Residual Heat Removal systems by a globe valve which provides a redundant function in this application.

s 87-0165 The Post Accident Sampling System Gas,Sampl'e Vial

< pressure instrumentation' indication did not meet j the Instrument Calibration Data Card specifications. This is attributed to improper mounting of the transducer into the sample tubing.

The equipment operating procedure is being revised to compute actual pressure from f.ndicated pressure until-the transducer can be moun'ted correctly.

87-0166 During the disassembly of a check valve in the Residual Heat Removal system, it was discovered that some parts were missing. Dae to the size end probable location of the missing parts, it was determined that no significant coolant . flow reduction or control rod motion interference would occur because of these missing parts.

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I Safety-Evaluation Description of Deficiency Report (DR) 87-0169 The required- outsido dimension: pipe checks were

. not perfortaed prior to establishing freeze. seals in support'of the repairs to the outer' elbow taps on the Reactor Recirculation Suctior Lines. The outnide dimension checks are to? ensure that -gross

' ove.lity deviations do not ocdur as- a reruit of freeze sealing. The -large pipe wall thickness combined with the small outside dimensions preclude these deviations. Additionally,-positive pressurei test results show. that no sirface cracking has occurred. '

A c$eck. valve in the piping =that plovides- Service

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~87-0171 Water to the Reactor. Auxiliaries Cooling System is

'being removed from the Inservice Test Program.

'The function of this check valve' is to prevent backflow when Service Water is " shutdown or l'solated f or maintenance. - There are other- valves upstream of this valve which provide redundant functions, 87-0174 During modifications:to the Undervemsel Equipment Handling Platform, 4 track clam,ps' were removed.

Only 1 was replaced. The platfois sits on a rail' f

that is rigidly supported by theLReactor Pressure ,,

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Vossel Pedestal Wall. This Saf.e*.y Evaluation.

included a II/I evaluation!which. determined that s the missing clamps do no t ' cor:stitute a seismic  ;.'

concern. l

~87-0175

'The'"B" Service Water Pump uncoupled test results  ;

show an indication' of a bowed pump shaft.

. After i the puup was re-coupled, vibration readings and '

pump flows were within the acceptance range.

Additionally, the "D" Service Water Pump would start on low "B" Service Water Pump Flow and provide an adequate back-up.

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4' 87-0177 The "C" Service Water Strainei Backwash hine has a pinhole 1tbk in it. The laak is downstIban of the strainer backwash valve and leaks? into the pump room. The leak presently has a clamp ca it to prevent further leakage. A deflector shield has been installed over the clamp to divert the flow in the event that the clamp fails.

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' Public Service Electric and Gas Company P.O. Box L Hancocks Bridge, New Jersey 08038 Hope Creek Operations -

J November 13, 1987 U. S. Nuclear Regulatory Commission Document Control Desk Washington,:DC 20555

Dear Sir:

HW.THLY OPERATING REPORT HOPE CREEK GENERATING' STATION UNIT 1 DOCKET NO. 50-354 In compliance with Section 6.9, Reporting Requirements for the Hope' Creek Technica1' Specifications, the operating statistics for October are being forwarded to you. In addition, the summary of changes, tests, and experiments for October 1987 are included r pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, q S. LaBruna ec General Manager - 1 Hope Creek Operations i RAR:t1b Attachment C Distribution I

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