ML20127L050
| ML20127L050 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/31/1992 |
| From: | Hagan J, Zabielski V Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9211230403 | |
| Download: ML20127L050 (17) | |
Text
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O PSIEG Put;bc Service Elecific ard Gas Compdny P O B04 236 Hancocks Brdge, New J0'sey 08038 Hope Creek Generating Station liovember 13, 1992 U.
S.
11uclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
M011TiiLY OPERATIl1G REPORT i
i llOPE CREEK GEllERATIOli STATIOli IlliIT 1 i
DOCKET 110. 50-354 In compliance with Section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating statistics for October are being forwarded to you along with l
the summary of changes, tests, and experiments for October 1992 persuant to the requirements of 10CFR50.59(b).
Sincer ly yours, W1 J
H gan j
Gene Manager -
l Hopo ek Operations l
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Distribution l-l l
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The Enercy Peonle
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IllDEX 11 UMBER SECTIOl{
OF PAGES Average Daily Unit Power Level.
1 Operating Data Report 2
Refueling Information.
1 Monthly Operating Summary.
1 Summary of Changes, Tests, and Experiments.
10
'+
...--_.m_
__._m AVERAGE DAILY UNIT POWER LEVEL f
I DOCKET NO.
50-354 7
UNIT Uppe Creek i
DATE 11/13/92
(
COMPLETED BY V. Zab.ielski TELEMIONE f609) 339-3506 i
MONTH October 1992 i
DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWa-Net)
(MWe-tiet) 1.
A 17.
2 2.
A 18.
2 3.
2 19.
D 4.
d 20.
A 5.
2 21.
A 6.
Q 22.
2 7.
D 23.
2 8.
A 24.
A 9.
2 25.
Q 10.
A 26.
n 11.
2
- 27.
- Q 12.
2 28.
A 13.
2 29.
A 14.
Q 30.
Q 15.
H 31.
p 16.
2 J
OPERATING DATA REPORT DOCKET NO.
50-354 UNIT Hope Creek DATE
_11/13/92
. l' COMPLETED BY V.
Zabielski
'l /pyh TELEPHONE (609) 339-3506 OPERATING STATUS 1.
Reporting Period October 1992 Gross Hours in Report Period 745 2.
Currently Authorized Power Level (MWt) 1222 Max. Depend. Capacity (MWo-Het) 1031 Design Electrical Rating (MWe-Het) 1957 3.
Power Level to which restricted (if any) (MWe-Net)
Mono 4.
Reasons for restriction (if any)
This Yr To ll2 Dill Date Cumulative 5.
No. of hours reactor was critical 0.0 5804.5 42,965.8 6.
Reactor reserve shutdown hours 0.0 222 0.0 7.
Hours generator on line 212 5742.0 42,316.6 8.
Unit reserve shutdown hours 2tQ 0.0 0.0 9.
Gross thermal energy generated 2
18,508,363 134,505,506 (MWH)
- 30. Gross electrical energy 0
6.145,590 AA2190,084 generated (MWH)
- 11. Het electrical energy generated 2
5.860.258 A.2.511,807
- 12. Reactor service factor 0.0 79.3 83.5
- 13. Reactor availability factor 0.0 12x1 83.5
- 14. Unit service factor 0.0 78.4 82.3
- 15. Unit availability factor 0.0 lata 82.3
- 16. Unit capacity factor (using MDC) 0.0 77.7 8222
- 17. Unit capacity factor 222 75.0 77.5 (Using Design MWe)
- 18. Unit forced outage rate 212 212 4.8
- 19. Shutdowns scheduled over next 6 months (type, date, & duration):
None
- 20. If shutdown at end of report period, estimated date of start-up:
11/10/92
_d
l OPERATING DATA REPORT UNIT SilOTDOWilS AND POWER REDUCTIONS DOCKET NO.
50-354 UNIT lione Creek DATE 11/13/92 COMPLETED BY V.
Zabielski TELEPilONE (609) 239-3506
'HONTli October 1992 METilOD OF SilUTTING DOWN Tile TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE No.
DATE S= SCHEDULED (liOUT
(1)
POWER (2)
ACTION / COMMENTS 8
10/1 S
745 C
4 Continuation of 4th Refueling _
Outage Summary
...J
4-REI'UELING INFORMATION
' *f
=
DOCKET ilo.
50-354
- 1 UNIT Hgpe Creek
, fi DATE
_13/13/92 C
COMP'*
30 BY S
Iollinasworth t
7if0 HONE (609) 339-1051 r~ttober 1992 "c
ling information has changed from last month:
s 4W' yo g
'.uled date for next refueling:
9/12/92 3.
be.
duled date f or resL.rt following refueling:
11/11/92 I
A.
Will Technical Specification changes or other license 3
amendments be required?
h.A h.3 No L
B.
Has the reload fuel design bean revios
- ty the Station I 7'"
Operating Review Committee?
Y ss X
No If no, wr i s
scheduled?
5.
5 :i eduled date(s) for submitting proposed licensing action:
UZA
=
s 6.
Important teenaing considerations associated with refueling:
Sarae fresh fuel as current cycle:
no new consideratioa-7.
Number of Fuel Annenblies:
A.
Incore 764 c
B.
In Spant Fuel Storage (prior to refueling) 760 C.
In Spent Fuel Storage (after refueling) 1G38 R.
Present licensed spent fuel storage capacity:
4006 Future spent fuel storace capacity:
4006 3.
Date of last ref ueling that can be discharged 11/4, 201Q to spent fuel pool assuming the present (EOC16) licensed capacity:
(does not allow for ful.1-core offload) ggy m '
. ~ _ _.
HOPE. CREEK' GENERATING: STATION MONTHLY OPERATING
SUMMARY
October 1992 The 4th-Refueling Outage began on September 12 and continued throughout the month of October.
As of October 31,-the unit was-planned to be back on line on November 10.
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SUMMARY
OF_: CHANGES,_l TESTS, hND EXPERIMENTS-4 l
FOR THE IlOPE CREEK GENERATING-_ STATION
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OCTOBER 1992 i
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The following items have been evaluated to detsrmine:
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 saalfunction of a different type than any evaluated previcuoly-in the safety analysis report may be created; or 3.
If the margin of safety as defined in the basis for any t2chnical specification is reducod.
The 10CFR50.59 Safety Evaluationsishowed-that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of 'he reactor These items did not chouge the-plant effluent re2 eases and did not alter-the existing.
environmental impact.
The 10CFR50.59 Safety Evaluations E
determined that no unreviewed safety or environmental questions are involved.
1 4
)
4 RCE Dgscrintion of Safety Evaluation 4EC-1010/05
_This DCP connected permanent power feeds from lighting panels to existing convenience lights and receptacles in the Nuclear Steam Supply System panels in the Main Control Room.
This DCP enhances the working environment for the performance of maintenance and sutveillance testing in either the main Ccatrol Room or the lower control equipment room.
The operability of the safety-related panels is not affected by this DCP.
The installation meets seismic and electrical separation criteria.
Therefore, this DCP does not involve any Unreviewed Safety Questiens.
4EC-1021/01 These DCPs removed some snubbers or replaced them 4EC-1021/02 with struts.
The affected snubbers were in the 4EC-1021/03 Main Steam Lines and their associated Safety Relief 4EC-1021/04 VLlve piping lines.
The ramoval of the snubbers and the-conversion to struts decreases the chance for the piping system to be in an unanalyzed condition due to snubber failure, incresses the reliability of safety related equipment, and reduces station man-rem.
All of the analyses to reduce the snubber.
population were performed per the ASME Boiler and Pressure Vessel Code requirements..
They meet the design intent of the UFSAR, including the postulated pipe break criteria.
Therefore, these DCPs do not involve any Unreviewed Safety Questions.
4EC-1021/05 These DCPs removed some snubbers or replacea them 4EC-3021/06 with struts.
The affected snubbers were in the-Reactor Recirculation _ lines, the Residual Heat Pemoval lines, and their related components and equipment.
The removal of the snubbers and the conversion to struts decreases the chance for the piping system to b3 in an unanalyzed condition due tr snubber failure, incr. roes the reliability of
,-fety related equipment, and reduces station man-
- tem, i
All of the analyses to reduce the snubber population were performed per the ASME Boiler and-Pressure Vessel Code requirements.
They meet the design intent of the UFSAR, including the postulated pipe break criteri a.
Therefore, these DCPs do not involve any Unrevi.ewed Safety Questions.
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4 QQE Descript' ion of Safety EvaluatioD 4EC-1034/01 These DCPs 2.odified the Emergehey Diesel Generator 4EC-Iv54/02 Starting Air skid supply lines from the Air ~ Dryers
-4EC-1054/03 to the Air Receiver Tanks.
The DCP includes the 4EC-1054/04 installatien of Liquid Drain Tanks to collect any-condensate in the supply lines and flush it through the floor drains.
The installation of new piping, fittings, and drain tanks will improve system performance and reliability because the condensate water will no longer collect in the receivers, thereby reducing corrosion.
The addition of the new valves will improve the availability of the_ Emergency Diesel Generators because the Starting Air Receiver Tanks can be fed from any compressor, allowing a compressor outage without affecting the operability of an Emergency Diesel Generator.
Therefore, these DCPs do not involve any Unroviewed Safety Questions.
4EC-3022/01 These DCPs installed guick disconnects for 4EC-3022/02 temperature switches in-the Emergency niesel 4EC-3022/03 Generators.
The installation of the quick 4EC-3072/04 disconnects will improve the maintainability of the temperature switches.
The temperature switches provide an alarm and have no control function.
There is no change in control circuitry or setpoints of any instruments that are important to safety.
Therefore, these DCPs do not involve any Uhreviewed Safety Questio.ns.
4EC-3104/01 This DCP provided Control Room nyerhead annunciation of Main Turbine and Feedwater Turbine sensor failure alarms that are fed from the sensors that input inte the two out of three trip logic for the tarbines.
The sensors currently feed'ccmputer points only.
The plant computer Control Room Integrated Display-System and the annunciator system are not Class lE and do not perform any safety-related functions.
Therefore, this DCP does not involve any Unreviewed-Safety Questions, i
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DCE -
Description of Safety Evaluation-4FC-3285/01 This DCP installed a double o-ring gask a-scal on the discharge flange of a pressure relict valve in the Core Spray system.
The installation of a double o-ring gask-o-seal disc on the discharge flange does not affect the function of the relief valve.
The disc seal is a passive component whose primary safety function is to mitigate the consequences of an accident by allowing Type B leak testing in lieu of Type A Integrated Leak Rate Testing following valve maintenance.
Therefore, this DCP does not involve any Unreviewed safety Questions.
4EC-3316/01 This DCP relocated the Radiation Monitoring console from near the Control Room back panel into the Control Room " horseshoe".
Moving the Radiation Monitoring console does not involve any functional change.
The console is not l
essential for safe shutdown of the plant and serves no active emergency function during an accident.
Therefore, this DCP does not involve any Unreviewed Safety Questions.
4EC-3343/01 These DCPs ramoved environmental seals on various 4EC-3343/03 transmitters and replaced them with environmentally qualified quick disconnecto.
The quick disconnects help to minimize stay time in the Radiological Control Area.
These DCPs did not change the design function or the qualification of the system.
Therefore, they did not involve any Unreviewed Safety Questions.
4EC-3374/01 This DCP replaced contrcl rods.
The nuclear life of the replaced control rods would have expired prior to the next refueling outage.
The nuclear and mechanical design of the new control reds is equal to or exceeds the design requirements of the original equipment; therefore, this DCP does not involve-any Unreviewed Safety Questions.
l
3 hE Description of Deficiency Report HTE 92-010 This DR addresses the installation of schedule 40 pipe instead of schedule 80 pips at several Station Service Water l' inch and 1 inch root valve lines.
Analysis of the lines indicate that the schedule 40 pipe could withstand the design basis earthquake.
Also,-if the schedule 40 pipe failed, the Safety Auxiliaries Cooling System Heat Exchanger Room is cqalpped with flooding alarms that indicate in the Control Room.
The pipe was replaced with schedule 80
-l pipe during the 4th Refueling Outage.
t Therefore, the use of schedulo 40 pipe did rot involve any Unreviewed Safety Questions.
HTE 92-160 This DR addresses Scram Outlet Valves in the Control Rod Drive system that were inadverte",tly rebuilt with the incorrect seat material.
This DR justifies the continued operation of these valves for two fuel cycles.
Analysis shows that the incorrect seat material would remain in acceptable condition for 144 scrams involvini extreme conditions.
Therefore, thid DR does not involve any Unreviewed Safety Questions.
HQA 92-203 This DR documents the failure a material supplier to adhere to the testing requirements of ASME Section II for a blind; flange installed in the
'A' Station Service Water Common Supply Header.
This DR allowed the flange to be used-as-is temporarily, but required its replacement pricr to the end of the 4th Refueling Outage.
The material tests required by ASME Section II have snown that the chemical composition of the blind flange is acceptable.
The deficiency identified concerr.s the f ailure to perform the mechanical tests on a test specimen that was heat treated along with the finished product.
A test sample was analyzed that came from the same heat code and heat number as the-finished product; however, this sampla was heat treated separately from the finished blind flange.
The test results fror this test. sample were acceptable.
Therefore, th's DR does not involve any Unreviewed Saf<ty Questions.
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i Procedure Revision Qgpcription of____ Sa fety Evaluation HC.IC-GP.SF-0001(Q)
This procedure revision provides guidance Rev 5
that allows Control Rod withdrawal when the core is off-loaded.
The guidance includes steps to jumper the Rod Position-Information system input to the Reactor Manual Control System and to jumper the low power setpoint contacts to the Rod ~ Sequence Control System.
The Reactor Manual Control System, Refueling Interlocks, and the Rod Position Indication System are part of the controls-that limit Control Rod motion during refueling.
The signal to the Safety Parameters Display System provides Control Rod information to the operator. _The margin of safety is to avoid an inadvertent criticality.
This procedure is used only when there is no fuel in the core; therefore this procedure does not involve-an Unreviewed Safety Question.
HC.MD-FR.KE-0003(Q)
This procedure revision includes a Rev 8
temporary change to allow the removal of the Reactor Pressure-Vessel Head Insulation package using slings and the Auxiliary Hook of the Polar Crane.
The UFSAR indicates that the Reactor Pressure Vessel Strongback is used to-lift the insulation package.
Failure of one of the support slings or one of the Polar Crapo redundant load wires would retain the load.
Ample capacity and redundancy were specified,.the consequences of a single failure using slings _and the Auxiliary Hook of the Polar Crane are the same as che previous system.
Therefore, this procedure does not involve an-Unreviewed Safcty Question.
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Procedure Revision Description of Safety Evaluation-HC.MD-GP.ZZ-0099(Z)
This new procedure eliminates the need for Rev 0
a temporary modification when either Service Air Compressor is out of service during electrical' bus and Turbine Auxiliaries Cooling System-outages.
It describes the steps-required to install and remove temporary compressors.-
The Instrument Air system has no safetyity.
related functions other than che integr of the piping through the containment-penetration.
Failure of the system will not cc.apromise any safety-related system or component or prevent a safe shutdown of the plant.
.Therefore, this' procedure does not involve an Unreviewed Safety Question.
HC.OP-GP.PB-0002(Q)
This new procedure establishes guidelines Rev 0
for the removal and return to service of the 4.16KV vital bus and provides direction for the installation of temporary power-when maintenance is to be performed during outages.
Malfunctions in the non-1E system (temporary power) le totally isolated from, and has no effect on, class 1E equi.pment that is vital to the safe shutdown of the plant.
This procedure will'only be implemented with the plant in Operating conditions-4, 5,
or *, when only two~of the four vital channels are required to be operable.
This procedure will only be implemented. hen channel
'B' is not required to be operable..Therefore, this procedure does not involve an Unreviewed Safety Question.
Procedure-Revisign Description of Safety Evaluation HC.OP-GP.PB-0003(Q)
This new-procedure establishes guidelines-Rev 0
for the-removal and return to service of the 4.16KV vital bus and provides direction for the installation of temporary power when maintenance is to ae performed during outages.
Malfunctions in the non-lE system (temporary power) is totally isolated from, y
and has no effect on, class 1E equipment that is vital to the safe shutdown of the plant-This procedure will-only be.
Implemented with the plant in Operating-Conditions 4, 5,
or *, when only.two of the four vital channels are required to be
- operable.
This, procedure will only be implemented when channel
'C' is not required to ce operable.
Therefore, this procedure does not involve an Unreviewed-Safety Question.
HC.SA-AP.ZZ-0049(Q)
This procedure revision Geletes Rev 6
HC.SA-Ap.ZP.-0049(Q), which has been superseded by NC.NA-AP.ZZ-0049(Q).
The-UFSAR states that station administrative =
orocedures provide station wi6e direction in areas that are common.to all station departments.
NC.NA-AP. ZZ-004 9 (Q) : includes the majority of the' administrative contrcls and department responsibilities previously contained in HC.SA-AP.ZZ-0049(Q).
The remaining ~ administrative controls and department; responsibilities have been included in department level procedurec.
Trensferring aoministrative~ controls and department responsibilities from one procedure to another does not impact the e
probability or consequences of any type of' accidrnt.
Therefore, this procedure does-not involve an Unreviewed Safety Question.
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Procedure Revisica Description of Safety Evalug jgul NC llA-AP.ZZ-0025(Q)
This procedure revision nodifies the-Rev i
storage arid use of combustibic materials and changes the method for completing the.
transient combustible evaluation.
This procedure does not changc the evaluations made of the fire protection equipment or change the criteria or assumptions used to develop _the Fire Hazard Analysis.
It introduces steps for the crevention of fire and minimizes the impact af fire on the station.
Therefore, this procedure does not invclve an Unreviewed Safety Question.
NC.NA-AP.ZZ-0050(Q)
This procedure revision enhances the Rev 1
Station Testing Program by the inclusion of guidance for the testing of major components after. painting of movable parts, linkage, shafte, and springs.
It also provides specific guidelines for testing ~of motor operated valves.
This revision does not change any previously analyzed testing requirement nor does it change any testing method.
Therefore, this procedure does not involve an Unreviewed Safety Question.
._