ML20077A468
| ML20077A468 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/31/1994 |
| From: | Lyons D Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20077A465 | List: |
| References | |
| NUDOCS 9411230223 | |
| Download: ML20077A468 (9) | |
Text
INDEX NUMBER SECTION OF PAGES Average Daily Unit Power Level.
1 Operating Data Report 3
Refueling Information 1
Monthly Operating Summary.
1 Summary of Changes, Tests, and Experiments.
3 9411230223 941115 PDR ADOCK 05000354 R
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OPERATING DATA REPORT i
DOCKET NO.
50-354 UNIT Hope Creeh DATE 11/08/94_
COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517 OPERATING STATUS 1.
Reporting Period October 1914 Gross Hours in Report Period 745 2.
Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity (MWe-Net) 1031 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 539.0 5648.9 58471.9 6.
Reactor reserve shutdown hours 0.0 0.0 0.0 7.
Hours generator on line 507.0 5506.9 57539.4 8.
Unit reserve shutdown hours 0.0 0.0 0.0 9.
Gross thermal energy generated 1638024 17653112 183616482 (MWH)
- 10. Gross electrical energy 549755 5848516 60812470 generated (MWH)
- 11. Net electrical energy generated 520804 5575901 58103585 (MWH)
- 12. Reactor service factor 72.3 77.4 84.8
- 13. Reactor availability factor 72.3 77.4 84.8
- 14. Unit service factor 68.1 75.5 83.4
- 15. Unit availability factor 68.1 75.5 83.4
- 16. Unit capacity factor (using MDC) 67.8 74.1 81.7
- 17. Unit capacity factor 65.5 71.6 79.0 (Using Design MWe)
- 18. Unit forced outage rate 31.9 8.5 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:
N/A
OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTICMC DOCKET NO.
50-354 UNIT Hope Creek DATE J_1.j_Q 8 / 94 COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517 MONTH October 1994 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.
10/2 F
238 A
3 AUTOMATIC SCRAM CAUSED BY DESIGN ERROR IN DIGITAL FEEDWATER CONTROL SYSTEM 2.
10/7 F
INCLUDED A
3 WHEN ATTEMPTING IN ABOVE RESTART OF UNIT AN EHC SYSTEM PROBLEM CAUSED A PROBLEM WITH THE TURBINE ROLL. THE OPERATOR CLOSED ALL TURBINE VALVES WHICH RESULTED IN AN AUTO REACTOR SCRAM.
3.
10/14 F
0 A
9 PARTIAL LOSS OF FEADWATER HEATING CAUSED BY A BLOWN FUSE.
EVENT WAS NOT
>20% BUT WAS SIGNIFACANT.
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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-354 UNIT liope Creek DATE 11/08/94 COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517 MONTH October 1994 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Not)
(MWe-Net) 1.
1047 17.
1062 2.
ERE 18, 1060 3.
2 19.
1055 4.
2 20.
1056 5.
2 21.
1059 6.
2 22.
1064 7.
2 23.
1045 8.
2 24, 1053 9.
2 25.
1064 10.
2 26.
1068 11.
2 27.
1064 12.
21 28.
1052 13.
970 29.
1073 14.
947 30.
1052 15.
1063 31.
1053 16.
1064
.~
)
.l 21 HOPE CREEK GENERATING STATICN MONTHLY OPERATING
SUMMARY
i October 1994 Hope Creek entered the month of October operating at 100% power.
A reactor SCRAM occurred on October 2 on a spurious signal and a design deficiency in the Digital Feedwater Control System.
The reactor was taken critical again on October 7th, but a problem occurred with the EHC system during the turbine roll.
The operator closed all Turbine Valves which caused an automatic reactor SCRAM.
The reactor was taken critical on October 11, 1994, and the unit was returned full power on October 12, 1994.
On October 14, 1994 there was a transient caused by a patial loss j
of feedwater heating.
Recovery was complete later that day.
As of October 31,1994 the plant has been on line for 20 consecutive-days.
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION 1
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l October 1994 The following items have been evaluated to determine:
1.
If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety i
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.
The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
These items did not change the plant effluent releases and did not alter the existing environmental impact.
The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
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4.
Tsmoorary Modification Summary 21 Safety Evaluation T-Mod 94-030: This' Temporary Modification installed an electrical jumper across the #2 Feedwater Heater Hi-Hi Level trip switches and installed a temporary keep fill line to the low side of the level transmitters.
This modification is performed due to spurious indications during power ascension and is removed at approximately 40 % Reactor Power.
This T-Mod does not increase i
the probability or the consequences of an accident listed in Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resulting in a turbine trip.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
Procedure Summary of Safety Evaluation HC.OP-SO.AB-0001(Oi: This Procedure for " Main Steam System Operation" is being changed to incorporate a line-up change.
The 1ABHV-F016 and 1ABHV-F019 valves where changed to be left in the closed position.
The UFSAR does not specifically address this procedure however section 13.5 requires specific procedures for operation of plant systems.
The change will affect only normal valve line-up of the system.
These valves are opened simply as an' operating convenience.
The function of these valves, to close during a LOCA, will not be affected as they will already be in their accident position.
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.
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Other Summary of Safety Evaluation j
UFSAR Chance Notice " Penetration Seals in Vestibule Room 3301A":
The Fire Hazards Analysis Table Section " Effects of Fire on Safe Shutdown and/or Radioactive Release" will be revised to provide information relevant to the construction and fire rating of the north wall and ceiling of vestibule room 3301A, which forms the south wall of corridor room 3301.
The revision provides clarifying information only.
An Engineering Evaluation concluded that the barrier integrity is not compromised.
The probability of occurrence of equipment malfunctions, fires, or fires causing equipment malfunctions due to this UFSAR change is not increased.
The barrier wall and ceiling will provide its intended protection with the existing penetration seals.
Therefore, this UFSAR Change Notice does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
=g.-
w-..
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e, SAB Chance Notice 94-34: UFSAR Section 9.2.8.2 states: "The service water in the RACS heat exchanger tube side is maintained at a higher pressure than the closed loop system in the heat exchanger shell side.
In the event of tube failure the service water leaks into the closed loop system to preclude the possibility of radioactive release to the environment in the unlikely condition that the RACS cooling loop becomes radioactive."
The referenced UFSAR statement is being revised to delete that statement.
Calculations were developed that determined the RACS and SSWS operating pressures at the heat exchanger.
These calculations have determined that the SWSS is not greater than the RACS operating pressure.
RACS is 46.0 PSIG operating pressure and SWSS is 25.4 PSIG operating pressure.
This change does not increase the probability of an accident as described in the SAR because the RAC, SWSS, and Process Radiation Monitoring systems function and operation do not change as a result of this change.
It will not affect the ability of the systems to detect radioactive leakage and the actions required to isolate the leak.
Radiological leak detection is provided for in the RACS system.
As stated in UFSAR Section 9.2.1.5, a Hi-Rad condition will alarm in the control room requiring operation action to isolate the RACS from SSW system.
Therefore, this UFSAR Change Notice 94-34 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 A-)-VARX-NSE-0727-1 "Eauivalent Replacement and Docupentation Update Generic Evaluation":
This Generic Safety Evaluation is prepared to address plant part and component changes due to obsolescence, unavailability or unreliability, and subsequent document update of the Hope Creek SAR.
The changes to the facility as described in the SAR to allow deletion of specific manufacturer, manufacture's model number, component characteristics or valve mark numbers presently identified in the SAR.
Although this document allows changes to all systems and components through out Hope Creek, the operational and/or design conditions will not be affected.
An Equivalency Evaluation will be performed as required to insure that all critical characteristics are met and that there will be no impact to plant operations as a result of the component or piece part change.
Therefore, this Generic Safety Evaluation does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
/
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e A-0-ZZ-NSE-0838-0 "Desian Chance Exclusion Zones gfo Various Artificial Island Buildinas":
This Safety Evaluation is to identify structures and justify their purpose as being outside the scope of the Nuclear Jurisdiction (NAP-8 Design Change Process does not apply to modifications performed in these structures) as described in the Exclusion Zone Technical Standard.
These structures are shown on drawing 252312 which is included in the Salem SAR.
The Structures include:
Carpenter Shop Change House 320 Combination Shop Fabrication Shop 800 Material Center with the following limitations:
a) Valve OKCV-516 and the piping outboard of the valve.
b) Valve OKDV-601 and piping outboard of the valve.
c) Plumbing Waste and Vent System as described on drawing P-10531-0 sheet 1.
d) Main Switchboard 00-B-540 as described on drawing E-18033-0 and Outdoor Substation No.12 as described on drawing FSK-Z-101-32.0 sheet 2.
Nuclear Department Administrative Building Nuclear Services Building Process Improvement Center i
Processing Center Sewage Treatment Center Weld Test Shop 901 Building I
902 Building These facilities and the equipment installed in these facilities are all separated from the power blocks and do not affect any i
operational parameters or system functions that could affect plant operations.
Therefore, this Safety Evaluation does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.