ML20151G255
| ML20151G255 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/31/1988 |
| From: | Jensen H, Zapolski M Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20151G258 | List: |
| References | |
| NUDOCS 8804190301 | |
| Download: ML20151G255 (19) | |
Text
- f '".': h _
11 0 '
.t ij' HOPE CREEK GENERATING STATION HONTHLY OPERATING
SUMMARY
MARCH 1988 Hope Creek entered the month of March in cold shutdown continuing its First Refueling Outage that commenced on February 13, 1988.
The refueling outage continued throughout the month.
R-009 RAR:tlb 4
)
I J
)
8804190301 880956 g(
PDR ADOCK 05000354 R
j-.; q.
.c.
['
AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.86-354 UNIT Hope Creek DATE 4/15/88 COMPLETED BY M.
Zacolski TELEPHONE (609) 339-3738 HONTH March 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (HWe-Net)
(MWe-Net) 1
- unit in refuelina 17 2
outaae for the entire 18 3
month - all values are 19 4
0.0 MWe-Net 20 5
21 l
6 22 7
23 8
24 9
25 10 26 11 27 12 28 13 29 14 15 16
4*l q ':
OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.86-354 UNIT Hoce Creek DATE 4/15/88 b
COMPLETED BY M. Zanolski REPORT HONTH March. 2988 TELEPHONE (609) 339-3738 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F FORCED DURATION REASON REDUCING CORRECTIVE ACTION /
NO.
DATE S SCHEDULED (HOURS)
(1)
POWER (2)
COMMENTS 2
3/1 S
'744.0 C
4 CONTINUATION OF REFUELING OUTAGE
SUMMARY
L-
I: - ; G' ;
OPERATING DATA REPORT DOCKET NO.86-354 UNIT Hooe Creek DATE 4/15/88 1/
COMPLETED BY H.
Jensen f
TELEPHONE (609) 339-5261 OPERATING STATUS 1.
REPORTING PERIOD March 1988 GROSS HOURS IN REPORTING PERIOD 744 2.
CURRENTLY AUTHORIZED POWER' LEVEL (MWt) 3293 MAX. DEPEND. CAPACITY (MWe-Net) 1067 (1)
DESIGN ELECTRICAL RATING (MWe-Net) 1067 (1) 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 0
1045.0 8903.1 6.
REACTOR RESERVE SHUTDOWN HOURS 0
0 0
7.
HOURS GENERATOR ON LINE O
1037.9 8783.0 8.
UNIT RESERVE SHUTDOWN HOURS 0
0 0
9.
GROSS THERMAL ENERGY GENERATED (HWH) 0 3,378,284
_27,186,852_
10.
GROSS ELECTRICAL dNERGY GENERATED (MWH) 0 1,133,985 9,045,682 11.
NET ELECTRICAL ENERGY GENERATED (MWH) 0 1,077,368 8,642,406 12.
REACTOR SERVICE FACTOR JtUA 47.8 79.3 13.
REACTOR AVAILABILITY FACTOR N/A 47.8 79.3 14.
UNIT SERVICE FACTOR N/A 47.5 78.2 15.
UNIT AVAILABILITY FACTOR N/A 47.8 79.3 16.
UNIT CAPACITY FACTOR (Usina Desian HDC) 0.0 46.2 72.1 17.
UNIT CAPACITY FACTOR (Usina Desian MWe) 0.0 46.4 72.0 18.
UNIT FORCED OUTAGE RATE O
O O
l 19.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, & DURATION):
None 20.
IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
4/15/88 (1) Auaust 1987 data is under management review.
l-i:
k.a; REFUELING INFORMATION COMPLETED BY:
Chris Brennan DOCKET NO.: 50-354 UNIT NAME: Hope Creek Unit 1 DATE: 4115/88 TELEPHONE: 3193 EXTENSION: N/A
. Month March 1988 1.
- Refuelina information has changed from last month:
First Report YES NO 2.
Scheduled date for next refueling:
11-04-89
-3.
Scheduled date for restart following refueling:
12-18-89 4.
A)
Will Technical Specification changes or other license amendments be required?
YES X
NO B)
Has the reload fuel design been reviewed by the Station Operating Review Committee?
YES NO X
If no, when is it scheduled?
6-18-89 5.
Scheduled date(s) for submittino proposed licensino action:
7-18-89 6.
Important licensino considerations associated with refuelino:
Information not presentiv available 7.
Number of Fuel Assemblier:
A) Incore 764 B) In Spent Fuel Storace 232 8.
Present licensed spent fuel storage capacity:
1108 Future spent fuel storace capacity:
4006 9.
Date of last refueling that can be discharged to spent fuel pool assumina the present licensed capacity:
12-18-89 f
I
~
- , - ~o ~ _ ; _
F a'
SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION =
MARCH 1988 a --
9 t
f i
l l
l l
l l
l L.
2-g 4
R The following Design Chance-Packages (DCPs) 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
~1ncreased; 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 as defined in the basis for any technical specification is reduced.
None of the DCPs created a new safety hazard 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.
L I
l l
l
, (; ~6*:;
QqE Descriction of Desian Chance Packace 7124-This DCP installed a flange assembly consisting of two flanges and a blank plate to isolate branch valves leadina from the High Pressure Coolant Injection System to the Residual Heat Removal Heat Exchangers.
These mechanical isolations will prevent steam from leakina into the Residual Heat Removal System via steam condensina lines.
7202 This DCP added orifice plates in the Residual Heat Removal and-Fuel Pool Heat Exchanger Loops to avoid cavitation of the butterfly valves.
(Note:
The Residual Heat Removal "A"
Loop will be completed at a later time.
The rest of the DCP has been installed).
The addition of the orifice plates will provide an improved desian for balancing the design flow rate in the heat exchancer flow paths.
4-HCO-86-0530 This DCP replaced under voltage relays in the Class lE 4.16kv busses with solid state relays.
The new relays will improve the trip time and implemant Technical Specification Amendment 7.
4-HME-86-0781 This DCP added a pressure switch to the discharge pipina of the Drywell Leak Detection -
Radiation Monitoring System to detect high pressure conditions when the containment isolation valves close.
The pressure switch will improve the long rance performance of the sample pump by shuttina it off so the motor will not overload when the containment isolation valves close.
4-EMP-86-0938 This DCP installed an air compressor and a
hich-pressure airline.
The air compressor will be used to test respirators and SCBA units.
It will also be used to fill the SCBA units.
4EC-1002/04 This
- DCP, relocated the Source Range Monitor / Intermediate Rance Monitor Preamp
- Panel, associated conduit and cable, and tubing, conduit, and cables for a
differential pressure transmitter.
This equipment interfered with the proposed Control Rod Drive Rebuild / Maintenance Facility.
4EC-1002/06 This DCP relocated junction boxes and associated devices under the vessel to make room for the Control Rod Drive Handlina Machine Platform.
This modification was in support of the Control Rod Drive Rebuild / Maintenance Facility.
4a
.y.
a s
QCE Descriotion of Desian Chance Packace 4EC-1006 This DCP modified the interlocks on Residual Heat Removal valves to prevent the inadvertent draining of the Reactor Vessel to the Suppression Pool through
.the Shutdown Cooling Lines.
The interlocks will prevent operators from opening the Residual Heat Removal Shutdown Cooling Valves,when the Residual Heat Removal ' System is.in the Shutdown-Coolina Mode.
4EC-1030/01 This DCP installed rigaing points for the removal of the "A"
and~"C" Core Spray Pumps and associated 1
floor pluas.
This will result in greater cost effectiveness when removina the pumps.
4EC-1030/04 This DCP installed rigaina points for the removal of the "B"
and "D" Core Spray Pumps and associated floor pluos.
This will result in creater cost effectiveness when removina the pumps.
^,,
4EC-1030/06 This DCP provided permanent support lugs welded to the Drywell Personnel Airlock Barrel in the Reactor Buildina.
It also provided a removable aluminum monorail bolted. to the support lugs to expedite the removal of equipment and tools for maintenance.
This will result.
in greater i
cost-effectiveness.
4EC-1030/08 This DCP provided permanant liftina arrangements for removina equipment and concrete floor plugs l
from the Reactor Buildina.
This will result in greater cost-effectiveness, r
4EC-1051 This DCP established continuous flow throuch the Off-Gas Radiation Monitoring System Sample Tank.
This was accomplished by installina a vacuum pump to induce flow and to provide flow to overcome the water seal in the Sample Discharoe Header to the condenser.
This DCP makes a
Terporary Modification permanent and continues to provide continuous flow throuch the Radiation Monitoring System
- Sampler, as required by Technical Specifications.
r I
L
.,. O ?,.
.V j
QQE Descriotion of Desian Chance Packace 4EC-10$7 This DCP added a_ control valve operator to each Chiller Condenser Turbine Auxiliaries Cooling System Outlet Butterfly Valves in the Turbine Building chillers.
A pressure controller will sense refrigerant pressure and provide a control signal to each valve operator to throttle Turbine Auxiliaries Coolina System coolina water through
~the chiller condensers.
This will prevent chiller trips due to seasonal temperature fluctuations.
4EC-1058/01 This DCP installed fiberglass re-inforced plastic enclosures around the Service Water Travelling Screen components.
The enclosures are utilized to eliminste water spray around the motor areas.
4EC-1075 This DCP relocated Heating, Ventilation, and Air conditionino Moisture Sensors and Transmitters from the Computer Room to the Control Room Return Air Ducts.
This modification will provide a more representative indication of the Control Room moisture level and improve the operability of the humidification /dehomidification systems.
4EC-1082/03 This DCP corrected discrepancies identified r
during the Control Room Desian Review Process.
The specific discrepancios corrected by this DCP r.re as follows: 1) tho alarms in the Control Room were divided into
- sections, with each section havina a
separats
- sound, 2) the secondary I
condsnsate feed pumps and valves were sequenced the same way as the primary condensate feed pumps e
and valves, 3) installed a more descriptive push button confiauration, and 4) corrected the Hydrogen-Oxygen Analyzer Recorder Scale.
4EC-1082/04 This DCP alioned the Intermediate Rance Mcnitor recorders with the Intermediate Range Monitor rance switches.
This discrepancy was identified durino the Control Room Jesign Review process and is part of a commitment to improve human factors in the Control Room.
4EC-1082/05 This DCP provided for indication of Reactor Vessel metal /flance temperatures in the Main Control Room.
This information is used to monitor thermal stress and was identified as a necessity during the Control Room Desion Review process.
I i
i t
f l
- J QqE Descrip_ tion _of_D_esign_ Change _ Package 4EC-1082/06 This DCP modified the Reactor Recirculation Flow Indicating Controller to make the indicated output "direct actina" with the valve position.
The previous configuration required the operator to increase the station output to close the recirculation flow control valve.
This discrepancy was identified during the Control Room Desian Review process and is part of a commitment to improve human factors in the Control Room.
4EC-1082/07 This DCP provided a warning licht in the Control Room to indicate when the Bailey Logic Cabinet has detected a "Containment Hich Pressure" and/or a
"Reactor Vessel Level Low Low" signal and latched in a
"half-trip" of the Primary Containment Isolation System Loss of Coolant Accident Level 2
Isolation Sicnal.
The need for this warnina light was identified durino the Control Room Design Review process and is part of a commitment to improve human factors in the Control Room.
4EC-1085 This DCP extended the ductwork in the Hioh Pressure Coolant Injection Pipe Chase Room to allow more efficient mixina of coolina air in the room.
The previous maximum temperatures exceeded the FSAR maximum temperature commitments and have accelerated the rate of equipment degradation.
This modification will rectify both of these problems.
4EC-1086 This DCP installed a demineralizer to the Reactor Auxiliaries Coolina System.
This demineralizer will maintain low conductivity water to provide an acceptable corrosion treatment as recommended by General Electric.
4EC-1087 This DCP installed a demineralizer to the Safety and Turbine Auxiliaries Cooling System.
This demineralizer will maintain low conductivity water to provide an acceptable corrosion treatment as recommended by General Electric.
4HC-0002 This DCP installed shutdown range reactor level indication on the Remote Shutdown Panel.
This will provide the operator with accurate level indication durina shutdown conditions and allows the operator to determine water level above
+60",
which is required for certain operational sequences.
- ,. sr g..
QqE Descriotion of Desian Chance Packace 4HC-0014 This DCP modified the Safety and Turbine Auxiliaries Cooling System Accumulatots by installina a floatina roof and a arating inside the accumulators and a diffuser at the Nitrogen Inlet Nozzle.
The floatina roof will act as a
barrier between the nitrogen and the water inside of the accumulators, minimizino the mixing of water and nitrocen.
The aratina will serve as a
vortex breaker and minimize turbulence inside of the accumulators.
The diffuser will diffuse the nitrogen jet impinaing upon the floating
- rod, therefore preventina an unbalanced force on the floatino roof.
4HC-0026 This DCP modified the Reactor Water Clean Up System to preclude the low flow lock-out of the domineralizers with subsequent backwashing and precoating by ensurina holding pump flow is established prior to trippina the recirculation pumps and closing the isolation valves.
This modification will reduce liquid radwaste inventory and processing costs and increase the reliability and availability of the Reactor Water Clean Up system.
- ',. O' /
The following Temporary Modification Requests (TMRs) have been evaluated to determinet 1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety l
previously evaluated in the safety analysis report may be increased; or l
l 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 as defined in the basis for any technical l
specification is reduced.
I None of the THRs created a new satety hazard to the plant nor did they affect the safe shutdown of the reactor.
These THRs did not change the plant effluent releases and did not alter the existing l
environmental impact.
The Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
i
Safety Evaluation Description of Temoorary Modification Recuest
( THR ).
88-0010 This THR provided a temporary power source to a
Non-1E 20kv Uninterruptable Power Supply Inverter during the maintenance outage of the "A"
4.16kv switchgear.
This modification is to be in use only when the "A"
4.16kv switchgear is inoperable and the plant is in operational condition 5.
88-0011 This THR provided a temporary power source to a
Non-1E Battery Cl.arger durina the maintenance outace of the "B"
4.16 kv switchaear.
This modification is to be in use only when the "B"
4.16 kv switchaear is inoperable and the plant is in operational condition 5.
88-0021 This THR blocked off the smoke detectors in the Reactor Buildina, elevation 256' 6" so that normal operation of the Polar Crane did not cause false nuisance alarms.
A Fire Watch was posted as a
compensatory measure durino this modification.
88-0022 This THR provided a 480 volt Non-1E temporary power source to a 125 volt DC battery charcer during the "A"
4.16 kv Class 1E bus outage.
This modification is to be used only when "A"
Channel ir inoperable and the plant is in operational conditions 4,5, and *.
88-0023 This THR provided a 480 volt Non-lE temporary power source to a 125 volt DC battery charger during the "B"
4.16 kV Class 1E bus outage.
This modification is to be used only when "B" Channel is inoperable and the plant is in operational conditions 4,5, and *.
88-0025 This THR provided a 480 volt Non-1E temporary power source to a 125 volt DC battery charcer durina the "D" 4.16 kV Class 1E bus outage.
This modification is to be used only when "D" Channel is inoperable and the plant is in operational conditions 4,5, and *.
88-0027 This THR provided a 480 volt Non-1E temporary power source to a 125 volt DC battery charcer during the "D"
4.16 kV Clr.ss lE bus outage.
This modification is to be used only when "D" Channel is inoperable and the plant is in operational conditions 4,5, and *.
':,. c.,.
Safety Evalyation Description of Temoorary Modification Reauest (THR) 88-0028 This THR provided a 480 volt Non-lE temporary power source to a 120 volt AC Class lE Public Address System Inverter during the "A" 4.16 kV
-Clads lE bus outaae.
This modification is to be used only when "A"
Channel is inoperable and the plant is in operational conditions 4,5.
and *.
88-0030 This THR provided a 480 volt Non-lE temporary power source to a 120 volt AC Nuclear Steam Supply System computer Inverter durina the "D"
4.16 kv Class lE bus outage.
This modification is to be used only when "D" Channel is inoperable and the plant is in operational conditions 4,5, and *.
88-0035 This THR installed blind flances for isolation to allow removal of a
valve in the "B"
Safety Auxiliaries Coolina System.
The valve was removed for repairs and re-installed.
At that time, the blind flances were removed.
88-0038 This THR provided temporary 120 volt AC power to a 24 volt DC Battery Charaer during the "B"
4.16 kv class lE bus outage.
This modification is to be used only when "B"
Channel is inoperable and the plant is in operational condition 4,5 or a.
88-0046 This THR provided a 480 volt Non-lE temporary power source to a Class lE Fuel Pool Cooling Pump durina the "A" 4.16 kV Class 1E bus outaae.
This modification is to be used only when "A" Channel is inoperable and the plant is in operational conditions 4,5, and *.
88-0057 This THR provided a 480 volt Non-lE temporary power source to a 120 volt AC distribution panel during the "A"
4.16 kv Class lE bus outaae.
This modification is to be used only when "A"
Channel is inoperable and the plant is in operational conditions 4.5, and *.
. s ' ;. ' :*
0 Safety Evaluation Des _c.riotion of Temoorary Modification Recuest (THR).
88-0065 The "A" Hydrogen Analyzer Calibration Gas Relief i
Valve leaks
- through, emptyin2 the bottle.
In i
order to prevent the loss of gas, the bottle could be valved closed.
This would cause annunciation in the Control Room.
This THR installed a jumper j
across the low pressure switch that causes the i
annunciation when the cas bottle is valved closed.
- ,.e',~
4 The following Deficiency Requests (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 d
t 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 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 chance the plant effluent releases and did not alter the existino environmental impact.
The Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
l l
\\
l I
1 1
P
7 y
oe Safety Evaluation Descr_iptio.n_of_Def_i_ciency Report (DR) 88-0031 An Emergency Diesel Combustion Air High Temperature Switch was discovered to have a nicked signal wire, which caused the switch to create a
ground.
The repair process for this nicked wire consists of butt spll.cino a new wire to the switch and applyinc the Raychem process.
This repair returns the switch to its original design.
88-0047 Durino the performance of a time response test on a Reactor Water Level Transmitter, the response time was outside the acceptable value.
The transmitter may be used
'as is" because its response time does not raise the response time for the loop above its Technical Specification Requirement.
88-0051 An ASME weld preheat of 69'F was used instead of the procedurally required 70*F.
The welds may be used "as is" because this did not violate the applicable ASME code section, 88-0052 An ASME weld preheat of 63*F was used instead of the procedurally required 70*F.
The welds may be used "as is" because this did not violate the applicable ASME code section.
88-0060 This DR deals with two instances of potentially lost parts.
These parts are a clove, or a wad of tape, or a polyethylene bag, and a pushbutton from a switch.
All of the potentially lost parts have been analyzed for the followino:
1) the potential for fuel bundle flow blockage and the subsequent fuel
- damage, 2) the potential for control rod interference, and 3) the potential for corrosion or other chemical reaction with reactor materials.
Safe reactor operation would not be compromised by the presence of the lost objects.
t