ML20066G586
| ML20066G586 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/31/1991 |
| From: | Hagan J, Zabielski V Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9102190187 | |
| Download: ML20066G586 (17) | |
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- Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038
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Hope Creek Generotmg Station '
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February.14_,
1991 1
'U.--S.. Nuclear Regulatory Commission Document Control Desk
-Washington, DC -20555
Dear:
Sir:
-MONTHLY-. OPERATING REPORT HOPE CREEK-GENERATION STATION UNIT 1 1 DOCKET NO.-50-354-Jn-compliance.with Section 6.9, Reporting Requirements for the J
i Hope'. Creek Technical Specifications, the operating statistics, for January are-being' forwarded to ~you.with the r
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'summaryL-of changes, tests, and experiments-1or January 1991-
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>pursuantito the' requirements of 10CFR50.59(b);.-
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- Sincerely;yours,
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.lHa an'___
h' Gene anager -
Hope' Creek Operations-W
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9102190:97 910131 PDR :ADOCK 05000354 P
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s INDEX NUMBER SECTION OF PAGES Average Daily Unit Power Level.
y operating Data Report 2
Refueling Information.
1 Monthly operating Summary.
1 Stimmary of Changes, Tests, and Experiments.
10
'1 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-354 UNIT HoDe Creek
.DATE 2/14/91 COMPLETED BY V.
Zabielski TELEPl!ONE -(609) 339-3J_Q1 MONT11 January 1991 DAY AVERAGE DAILY. POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(LNo-Net) 1.-
Q 17.
2
.3.
A 18.
2 3.
2' 19.
2 4h 2
20, 2
5.-
2 21.
2
'6.
2 22.
D 7.-
D 23.
?
8.
2 24.
p
'9.
.2-25.
p
- 10..
D-26.
A 111..
2 27.
2 13.
H-28.
Q 13. -
,D 29.
A
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(14.
A 30.
9 15...
9 31.
A 16.
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OPERATING DATA REPORT DOCKET NO.
50-354
- UNIT _ }ipp_p Creek DATE 2/14/91 -
COMPLETED BY L. Zabielski j
TELEPHONE.L6 09 ) 339-3506 OPERATING: STATUS 1.
- Reporting Period January 1991 Gross Hours in Report Period 2_4_4.
- 2.
Currently. Authorizud Power Level (MWt) 3293 Max. Depend. Capacity-1031 Design Electrical Ratin(MWe-Net) g (MWe-Net) 1067 i
'3.-
Power? Level to which restricted (if any)- (MWe-Net)
None 14. - Reasons for restriction ~(if any)-
This Yr To Month Date _
Cumulative i
- 5.-
No.'of hours reactor was critical b_Q L_Q 29.781.5 6.
- Reactor reserve shutdown hours-M L_Q L_Q
'7_. '
Hours generator!on line L_Q :
L.Q 29.293.1 1
8.
Unittreserve shutdown hours-M L_Q La 9..
Gross thermal energy generated-A A
92.542.408' 3
(MWH) 110. Gross' electrical; energy' A
2 30.621.673 generated - (MWH).:
11.' Net; electrical energy generated a
. _Q -
29.248.396 (MWH)
- 13. ~ Reactor service f actor _
L_Q 0.0 82.5
' !13. Reactor avai-lability factor =
L_Q L_Q -
82.5 J14. Unit ~ s'ery;i'ce - factor LH
.LS 81.2 (15 yUnitfavailability factor:
L_Q L_Q al.d.
4 16;. Unit: capacity factor. (using MDC)
L.Q LQ 7 8, ()
il7.5.Unittcapacity factor L_Q L_q 75.9
'(Using Design-MWe)
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n L18; Unitiforcedooutage rate, LA L_Q
=L1 J19. - Shutdowns scheduled sover next c 6 monthe :(type, date, & duration):
. - None-l20.-IfUshutdown;at end=of-report period,. estimated date ofi start-up:,
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-'2/13/91 i
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OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-354 UNIT Hope Creek.
DATE 2/14/91 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH -January 1991 METHOD OF SHUTTING DOWN THE-TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.
DATE S= SCHEDULED (HOURS)
(1=)
POWER (2)
ACTION / COMMENTS 1
1/1 S
744 C
4 3rd.Refu911ng Outage Summary i
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REFUELING INFORMATION DOCKET NO.
50-354 UNIT Hope Creek l
t DATE 2/14/9:
COMPLETED BY S. Holl:.nasworth TELEPHONE (609) 339>1051 MONTH January 1991 1.
Refueling informacion has changed from last month:
Yes No X
2.
Scheduled date for next refueling:
12/26/90 3.
Scheduled date for restart following refueling:
p2/13/91 4.
A.
Will Technical Specification changes or ott,r license E
amendments be required?
Yes No' X B.
Has the reload fuel design been reviewed by the Station Operating Review Committee?
Yes=
No X
If no, when is it scheduled?
not currentiv scheduled S.-
Scheduled date(s) for submitting proposed licensing action:
HZA 6.
Important licensing considerations associated with refueling:
- Amendment 34 to the Hope Creek Tech Spocs allows the cycle specific operating limits to be incorporated into the CORE OPERATING' LIMITS REPORT; a submittal is therefore not i
. required.
1-7.-
Number:of Fuel Assemblies:
'A.-
Incore 764 B.
In Spent Fuel Storage (prior to refueling)
Agg C.-
In Spent Fuel Storage (after refueling) 760 y
8.
Present licensed spent fuel storage capacity'.
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Future spent fuel storage capacity:
4006 9.-
Date of last refueling that can be dischr.rged J21v R22_.2007 l
to spent fuel pool. assuming the present licensed capacity:
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HOPE CREEK-GENERATING STATION MONTHLY OPERATING
SUMMARY
JANUARY 1991 At the beginning of February, Hope Creek remained shutdown for the third refueling outage.
The outage continued throughout the-month.
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS 4 -
FOR THE HOPE CREEK GENERATING STATION JANUARY 1991 i
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- The following Design Change Packages (DCP's) have been evaluated to determine:
1.
If the probability of occurrence or the ennsequences of an accident or malfunction of equipment iraport0nt to safety previously evaluated in the safety analysis report may be increased; or 2.
If a possibility for an accident or malfunction of.a d1Lforent type than any evaluated previous?.y in-the safety analysis report may be created; or 3.
If the margin of safety as definad in the basis for any technical specification is reduced.
The DCP's did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
The DCP's 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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i HRE Description of Desian Chance Packacq 4EC-3008 This DCP added rigging attachment points to allow for easier removal and replacement of valves and actuators.
The valves and actuators addressed in i
this DCP are in the Containment Atmosphere Control System.
4EC-3039 This DCP upgraded the closed circuit television system that is used to observe the packaging and drum handling portions of the Solid Radwaste System.
The DCP will enhance operation of this system through additional / upgraded cameras.
4FC-3130/01 This DCP installed two monorails to service the Reactor Water Cleanup System pumps and motors in the Reactor Building Reactor Water Cleanup Recirculation Pump Rooms.
The monorails will enhance pump maintenance by providing tha capabilit/ to lift either the pumps or the motors.
4EC-3148 This DCP installed a tide flex check valve at the end of each of the two Storm Drainage Cutfall pipes.
The-check valves will prevent river water and sediment from flowing into the Storm Drainage System.
4EC-3149/02 This DCP added rigging attachment points to' allow for easier removal and replacement of valves and actuators.
The valves and actuators addressed.in this DCP are in the Torus Water Cleanup, Safety Auxiliaries Cooling, and Containment Atmosphere Control Systems.
4HC-0195/04 This DCP replaced the inboard and c.tbohrd Containment Isolation Valves for the process sampling line off of the Nuclear Boiler System.
The replacement valves have a longer life expectancy.
4HC-0212/02 This DCP upgraded'the "B" Chilled Water Pump by replacing the motor and impeller.
This upgrade will allow.the Chilled Water System.to operate with 2 pumps' running and the other ln standby.
4HC-0212/03 This DCP upgraded the "C" Chilled Water Pump by replacing the motor and impeller.
This upgrade will allow the Chilled Water System to operate with 2 pumps running and the other in standby.
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DEE Description (if Desian Chance Packaae 4HC-0212/04 This DCP installed two stainless steel pipe tie-ins to the Chilled Water System, one on the suction header of the Chilled Water Pumps und one on the discharge header.
These tie-ins support the future inctallation or a mechanical filtering unit and a mixed bed demineralizer to cleanse the system of solublo impurities that reduce performance and equipment life.
4HC-0238/01 This DCP modified the Safety and Turbine Auxiliaries Cooling System piping and instrument tubing fra the flow measurement used to monitor the Sa/et" Aux.liaries Cooling System to Turbine humilIaries Cooling System flow.
This DCP will reduce the flow measurement errors that have resulted in spurious actuationc.
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1 The following Temparary Modification Requests (TMR's). lave been evaluated to determine:
1.
If'the probability of occurrence or the concequerces of an accident or malfunction of equipment important t's safety previously evaluated in the eafety 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.
Tho'TMR's did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
The TMR's 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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THE Descrintion of Temocrary Modification Reauest 90-053 This TMR installed a power cable to provide a temporary power source to the
'A' Reactor Protection System Power Distribution Panel'during the maintenance outage on the 4160 Volt Turbine Building Switchgear.
90-0./
This TMR installed a-power cable to provide W
temporary power to a 120 Volt Miscellaneous Instrument Power Supply during the maintenance outage on the
'A' 4160 Volt-Class 1E Switchgear.90-058 This TMR installed a power cable to provide temporary power to the Public Address System Power Supply during the maintenance outaan on the
'A' 4160 Volt Class 1E Switchgear.90-062 This TMR installed a power cable to provide temporary power to a 125 Volt Class 1E Battery Charger during the maintenance outage on the
'A' 4160 Volt Class 1E Switchgear.90-063 This TMR installed a power cable to provide
= temporary power to the
'D' class 1E 120VAC NSSS Computer Power Supply during the maintenance outage on the
'D' 4160 Volt Class 14 Switchgear.90-064 This TMR installed a power cable to pro,'ide temporary power to a 24 Volt Battery Charger during the maintenance outage on the
'D' 4 1 (. ) Volt Class 1E Switchgear.
.90-065 This TMR installed a power cable to provide temporary power to a
'D' Class 1E 125 Volt Batter:
Cherger during the maintenance outage on the
'D' 4160 Volt Class lE Switchgear.90-066 This TMR installed a power cable to provide temporary power to a
'D' Class 1E 125 Volt Battery Charger'during the maintenance outage on the
'D' 4160 Volt Class 1E Switchgear.-
.91-001 This TMR added a connection to a temporary Air compressor in the discharge 1ine of the Service Air Compressor Aftercooler.
The temporary compressors will be used to supply Service Air while the Service Air Compressors are out of service for a maintenance outage.
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- The following Deficiency! Reports-(DR's) have-been evaluated to det0rmine::
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1.
If the: probability of; occurrence cur the consequences of an accidant or:. malfunction of equipment important to safety
-previously. evaluated =in the safety analysis report may be increased; -cur
-1 2.
If a possibility.for an accident or malfunction ~of a different type than,any evaluated previously in the safety-analysis 1
report may be created; or--
3.>
If the margin of safety as defined in the basis for any technical specification is reduced.
The: DR8 s? did-not croatei a new safety hazard to -the plant nor did -
Lthey. affect the-safia shutdown of the reactor.
The DR's=did not change-the plant effluent releases and did not alter the existing-environmental impact.
The Safety Evaluations determined that no unreviewed sefety or: environmental questions l are involved.
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1 DE Descriotion of Deficiency Report--
t HTE-90-082 This DR addresses the overtorquing of Fuel Channel Fastener bolts.
The procedure states that the-j bolts should be torqued to 75 i 5 inch-pounds.
The torque wrench that was used on 58 fuel bundles was
.found to exhibit a torgue of 81.37 inch-pounds.
The DR authorizes'the ruel bundles to be "used-as-is".
HMD-91-004 These three DR's were evaluated together.--They-HMT-91-016 Lddress wold cracks on a hanger in the HMD-91-021
- Recirculation System and linear indications on the.
outside surface of the weld between the
..ecirculation Loop Discharge pips'and the Residual Heat Removal returnitee in both loops..The--Safety Evaluations associated-with these DR's state-that the Hope Creek principal safety barriers were not seriously degraded and that the Recirculation.
. System may be "used-as-is" while:the plant.is shutdown and' operating in-Residual Heat Removal J
- HMD-91-027 This'DR addresses a1 Main Steam Drain Valve that has a seat contact area of less=than 100%.
The valve cannot be replaced 5at this time because of the unavailability =ofta spare valve.
The drain valve may:be "used-as-is".because=there is another drain valve and a pipe cap;in series, which meet the intent of the-double-valve design..
RNR MC91-0032 This Receiving Nonconformance Report identifies a Enonconformance with purchase specifications.
The affected component-isyc. butterfly valve.for;the r
Service Water' System..-The purchase order specified.
a-leakage rate of O cc/ min, which cannot be-
-achieved by thin vaAve. -The. valve may be "used-as-is" -becaus e the leakt ge rate is minimal..
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The following procedure revisions have been evaluated to i
determine:.
_ 1-If the probability of occurrence or the? consequences of an accident-or: malfunction of equipment important to safety rpreviously 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 q
report.may'be' created;-or 3.
IfTthe, margin of safety as defined in the' basis for any technical.epecification is reduced.
The procedure revisions did not-create a new safety hazard to the
- plant nor did'they affect the safe shutdown of the reactor.
.The procedure revisions 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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i-Procedure Revision Descrintion of Procedure Revision HC.OP-IS.LC-0002(Q)
This Safety Evaluation addresses the use of Rev. 7 temporary Measurement & Test Equipment instead of using permanently installed instrumentation as stated in the UFSAR.
The use of the temporary Measurement & Test Equipment is consistent with ASME Section XI and Hope Creek's Inservice Test program.
HC.OP-SO.SE-0001(Q)
This procedure revision adds a section to Rev. 2
' defeat the downscale Rod Blocks when all fuel is removed from the Reactor Vessel.
Rod Blocks are not required when there is no fuel in the Reactor Vessel.
HC.OP-SO.SF-0001(Q)
Thin procedure revision allows the Scram Rev. 2 Discharge Volume High Level Rod Block to be bypassed when all fuel is removed from the Reactor Vessel.
Rod Blocks are not required when there is no fuel in the Reactor Vessel.
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