ML20080C609
| ML20080C609 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/30/1994 |
| From: | Hovey R, Lyons D Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9412190087 | |
| Download: ML20080C609 (10) | |
Text
.
-l O PSIEG
- PublicServicokloctric and Gas Company P,0, Box '236 Hancocks Bridge, New Jersey 08038 J
Hope Creek Generating Station
-, December 15, 1994' U.
S.
Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
MONTHLY OPERATING REPORT HOPE CREEK GENERATION STATION UNIT 1 DOCKET NO. 50-354 In compliance with Section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating statistics for November are being forwarded to you with the summary of changes, tests, and experiments that were implemented during' November 1994 pursuant to the requirements of 10CFR50.59(b).
Sincerely yours, R.d.
General Manager -
M Hope Creek Operations R
S:JC Attachments C
Distribution i
I The Enerov People 9412190087 941130 PDR ADOCK 05000354
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.'J 3-INDEX l
H 4
NUMBER
-SECTION OF PAGES Average Daily Unit Power Level.
1 Operating Data' Report.
..c.
3
-Refueling Information...............-.
l' Monthly Operating Summary.
'1.
Summary of Changes, Tests, and Experiments.
2 1
j
OPERATING DATA REPORT DOCKET NO.
50-354 UNIT Hope Creek DATE 12/06/94 COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517
' OPERATING STATUS
.1.
Reporting Period November 1994 Gross Hours in Report Period 21Q' 3.
Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity 1031 Design Electrical Ratin(MWe-Net) g (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 720.0 6368.9 59191.9 6.
~ Reactor reserve shutdown hours gzQ 0.0 AzQ 7.
Hours generator on line 720.0 6226.9 58259.4 8.
Unit reserve shutdown hours AzQ 0.0 gzQ 9.
Gross thermal energy generated 2367549 20020661 185984030 (MWH)
- 10. Gross electrical energy 795980 6644497 61608451 generated (MWH)
- 11. Net electrical energy generated 763549 6339451 58867134 (MWH)
- 12. Reactor service factor 100.0 79.5 85.0
- 13. Reactor availability factor 100.0 79.5 85,0
- 14. Unit service factor 100.0 77.7 83.6
- 15. Unit availability factor 100.0 77.7 83,6
- 16. Unit capacity factor (using MDC) 102.9 76.7 82.0
- 17. Unit capacity factor 99.4 74.1 22x2 (Using Design MWe)
- 18. Unit forced outage rate AzQ IzQ 4.0
- 19. Shutdowns scheduled over next 6 months (type, date, & duration):
None
- 20. If shutdown at end of report period, estimated date of start-up:
N/A 6y4"-
e+--wt' W:
e-.=
A m---
OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.
S'0-354 UNIT HoDe Creek-DATE 12/06/94 COMPLETED BY D.
W.
Lyons TELEPHONE.
(609) 339-35,12 MONTH November 1994 e
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.
NONE l
1
1 f-t
' AVERAGE DAILY UNIT POWER LEVEL-DOCKET NO.,
50-354-UNIT.__ Hone Creek
-DATE 12/06/94 L
COMPLETED BY D.
W.
Lvons TELEPHONE (609) 339-3517 i
l MONTH November 1994
' DAY AVERAGE DAILY POWER LEVEL.
DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
-(MWe-Net) 1.
1057.
17.
1061 2.
1064 18.
1060 3.
1068 19.
1060 4.
1055 20.
1052 5.
1212 21.
1057 6.
lAll 22.
1056 7.
1061 23.
1066 8.
1060 24.
1070 9.
1054 25, 1064 10.
1063 26..
1065 11.
1072 27.
1061 12.
1066 28.
1044 13.
1066 29.
1QJ6_q 14.
1051 30.
1075 15.
1Q62 31.
D/_a 16.
1068
1,1, I
'e s
REFUELING INFORMATION
. DOCKET NO.
50-354 UNIT Hone Creek.1 DATE 12/06/94 COMPLETED BY R.
Schmidt TELEPHONE (609) 339-3740
- MONTH LDecember 1994
- l'.1 Refueling information has changed from last month:
Yes-No X
2.
Scheduled date for next refueling:
2116/95 (Under Review)
-3.
Scheduled date for restart following refueling:
10/31/95 4 '..
A.
Will Technical Specification changes or other license amendments be required?
Yes No X B.
Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee?
Yes No X
If no, when is it scheduled?
Auaust 28. 1995 5.
Scheduled date(s) for submitting proposed licensing' action:
Hgt reauired.
6.
Important' licensing considerations associated with refueling:
8.!.h 7.
Number of Fuel Assemblies:
A.
Incore 2fA-B.
In Spent Fuel Storage-(prior to refueling) 111.q C.
In Spent Fuel Storage (after refueling) 1472 8.
Present licensed spent fuel storage capacity:
4006 Future spent fuel storage capacity:
AQOg 9.
Date of last refueling that can be discharged S/3/2006 to spent' fuel pool assuming the present (EOC13) licensed capacity:
(Does allow for full-core offload)
(Assumes 244 bundle reloads every 18 months until then)
(Does nqt allow for smaller reloads due to improved fuel)
~
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'j
. o HOPE CREEK GENERATING STATION MONTHLY OPERATING'
SUMMARY
'r November 1994 Hope Creek' entered the month of November operating'at 100%'for.the month.- The. unit operated at full power.throu without.any major power' reductions or scrams.gh out the monthAs_of November 30, 1994 the unit has been.on-line for 50 days..
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION November 1994 f
The following items have been evaluated to determine:-
-5 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 - 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.
l The 10CFR50.59 Safety Evaluations showed'that these items did not create a new safety hazard to the plant nor did they' affect the i
safe shutdo.n of the reactor.
These items did not' change the.
1 plant affluent releases and did not alter the existing environmental impact.
The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
t
.t f
1 a
m Desian Chances Summary 21 Safety Evaluation 4$C-00032 Pkg 3 & 4:
These Design Change Packages included the replacement of the non safety Reactor Protection System (RPS) alternate feed BUS transformers with "ISOREG" 25 KVA Regulating Transformers.
The "ISOREG" Regulating Transformer will improve the stability of the 120VAC power supplied to the Electrical Protection Assemblies (EPA's) and ri '.uce noise in the system.
These installations do involve changing the EPA's existing UV/OV setpoints.
The RPS alternate feed power with "ISOREG" Regulating Transformer will improve the reliability of the 120VAC power supplied to the EPA's and prevent their tripping due to spurious undervoltage, overvoltage, and frequency swings.
The system noise is also reduced.
Therefore there are no credible failure modes created as a result of these change packages.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
Procedure Summary 91 Safety Evaluation Procedure NC.NA-AP.ZZ-0065(Z) Rev 2:
This procedure which governed Nuclear Department Resource Allocation has been deleted as a Nuclear Department Administrative Procedure.
A new Nuclear Department Business Procedure was created to cover this information.
The Operational Quality Assurance Program does not apply to this procedure.
Although this procedure is briefly mentioned in the UFSAR (Sect 13.5.1) it performs no function relative to the safe operation of Hope Creek Station.
Therefore, this Procedure revision does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
Other gummary 91 Ea f ety Evaluation UFSAR Change 94-43:
These UFSAR Changes are the result of the UFSAR Change 94-25 wherein the terms Type "A" and Containment Intergrated Leakage Rate Test (CILRT) were redefined and removes the ambiguity as to whether these actions are acceptable.
This change allows the Type "A" test to be completed in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
It provides for leakage monitoring of the CRD lines anytime during the CILRT as long as the reactor vessel and non-seismic portion of the system is vented and containment is pressurized, not just during the Type "A"
test.
It identifies the purpose of the Type "A" test as the measurement of air leakage from the containment.
It requires the Type "C" testing of penetrations isolated from the Type "A" test be used to adjust the Type "A" measured leakage rate.
It removes the requirement to operate drywell chilled water during the Type "A" test, but requires that it be maintained available if needed which will allow more options to maintain a stable containment atmosphere.
The dFSAR changes do not revise limits or acceptance criteria contained in the Technical Specifications (TS).
Margins of safety as defined in the basis for TS's remained unchanged.
The Type "A"
test and supplemental test acceptance criteria remained unchanged.
The Wpe "A" test will continue to ensure that the primary containment atmosphere (air) leakage rate does not exceed 0.375 percent per day which provides 6 margin for operational degradation.
Therefore, this UFSAR Change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
Safety Evaluation 1C-8-501:
"C" Traveling Screen was replaced with a refurbished screen.
To replace this screen a 3" ASME Class 3 stainless steel line was cut to facilitate the screen removal.
The 3" line was rewelded and a final weld examination, Liquid Penetrant Test (PT), was performed and resulted in two indications.
Both indications were in the reinforcement (crown) portion of the weld.
Both indications were removed by filing and a PT was performed of the excavated site by the welder.
The PT on the excavated area was satisfactory and the welder replaced the reinforcement portion of the weld.
A final weld PT was performed by a qualified individual with satisfactory results.
All the steps of this weld repair followed the rules set forth by the ASME Code except the PT on the excavated area which was performed by the welder who was not qualified to ASME Section XI IWA-2300 requirements.
This screen was placed inservice for approximately 7 months until the QA engineer reviewed the work package and identified a problem with the qualifications of the individual performing the excavation PT.
The weld was re-excavated and satisfactorily repaired.
UFSAR Chapter 15 accident analysis discusses total failure of this one wold on one of three spray wash pipes of the "C" traveling screen could cause river debris to be carried over the screen and the "C" Service Water Pump suction.
The debris would travel through the pump and impinge on the "C" strainer mesh.
The "C"
strainer is designed to remove debris from river water.
If the strainer becomes overloaded with debris, control. room indications will alert the operators by overhead annunciator and by CRIDS points which would indicate an increase differential pressure across the service water strainer.
Therefore, this UFSAR Change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.