ML20140H676
| ML20140H676 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/31/1997 |
| From: | Kepley L, Todd F Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20140H090 | List: |
| References | |
| NUDOCS 9706180408 | |
| Download: ML20140H676 (11) | |
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INDEX NUMBER SECTION OF PAGES-Operating Data Report................................
2 Average Daily Unit Power Level.......................
1 Refueling Information................................
1 Monthly Operating Summary............................
1 Summary of Changes, Tests, and Experiments...........
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9706180408 970613 PDR ADOCK 05000354 R
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i DOCKET No.: 50-354 UNIT: Hope Creek DATE: 06/05/97 COMPLETED BY:
F. Todd TELEPHONE: (609) 339-1316 OPERATING DATA REPORT OPERATING STATUS
- 1. Reporting Period May 1997 Gross Hours in Report Period 744
- 2. Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity (MWe-Net) 1031 l
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 Cumulative Month Date 5.
No. of hours reactor was 744.0 3623 77346
)
critical 6.
Reactor reserve shutdown 0.0 0.0 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 7.
Hours generator on line 744.0 3623 76183 8.
Unit reserve shutdown hours 0.0 0.0 0.0 9.
Gross thermal energy 2424650 11670194 243538441 generated (MWH)
- 10. Gross electrical energy 803760 3940380 80834593 generated (MWH)
- 11. Net electrical energy 771635 3786611 77266007 generated (MWH)
- 12. Reactor service factor 100.0 100.0 84.4
- 13. Reactor availability factor 100.0 100.0 84.4
- 14. Unit service factor 100.0 100.0 83.2
- 15. Unit availability factor 100.0 100.0 83.2
- 16. Unit capacity factor (using 100.6 101.4 81.8 MDC)
- 17. Unit capacity factor (using 97.2 98.0 79.1 Design MWe)
- 18. Unit forced outage rate 0.0 0.0 4.2 19.
Shutdowns scheduled over next 6 months (type, date, &
duration):
j Refueling Outage, September 13, 1997, 50 days l
20.
If shutdown at end of report period, estimated date of l
start-up:
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l DOCKET NO.: 50-354
[
UNIT: Hope Creek l'
DATE: 06/05/97 l
COMPLETED BY:
F.
Todd
]
. TELEPHONE: (609) 339-1316 j
l OPERATING DATA REPORT i
UNIT SHUTDOWNS AND POWER REDUCTIONS MONTH MAY 1997 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.
DATE S= SCHEDULED (HOURS)
(1)
POWER (2)
ACTION / COMMENTS i
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DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 06/05/97 COMPLETED BY:
F.
Todd TELEPHONE: (609) 339-1316 AVERAGE DAILY UNIT POWER LEVEL MONTH MAY 1997 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 1096 17 1093 2
1090 18 1095 3
1088 19 1080 4
1093 20 1086 5
1097 21 1094 6
1089 22 1094 7
1094 23 1093 8
1095 24 1088 9
1094 25 1079 10 1095 26 1089 11 1075 27 975 12 1089 28 1083 13 1094 29 1092 14 1095 30 1087 15 1088 31 893 16 1095
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DOCKET NO..
50-354 l
UNIT: Hope Creek DATE: 06/05/97 COMPLETED BY:
L.
Kepley TELEPHONE: (609) 339-1106 REFUELING INFORMATION MONTH MAY 1997 1.
Refueling information has changed from last month:
Yes X
No 2.
Scheduled date for next refueling (RF07) : 9/13/9 3.
Scheduled date for restart following refueling:
11/12/97 i
4A. Will Technical Specification changes or other license amendments be required?
Yes X
No B.
Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee (SORC) ?
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Yes No X
If no, when is it scheduled?
To Be Determined for Cycle 8 COLR 5.
Scheduled date(s) for submitting proposed licensi j action:
A License Change Request will be submitted to the NRC on or before June 30, 1997.
6.
Important licensing considerations associated with refueling:
The above required LCR concerns the Technical Specification Safety Limit Minimum Critical Power Ratio for Cycle 8.
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7 Number of Fuel Assemblies:
A.
Incore 764 B.
In Spent Fuel Storage 1472 8.
Present licensed spent fuel storage capacity: 4006 Future spent fuel storage capacity:
4006 9.
Date of last refueling that can be discharged 5/3/2006 to spent fuel pool assuming the present licensed capacity:
(EOC13)
(Does allow for full-core off-load) l (Assumes 244 bundle reloads every 18 months until then) l (Does not allow for smaller reloads due to improved fuel) l l
l DOCKET No,: 50-354 UNIT: Hope Creek DATE: 06/05/97 COMPLETED BY:
F. Todd TELEPHONE: (609) 339-1316 MONTHLY OPERATING
SUMMARY
MONTH MAY 1997 The Hope Creek Generating Station remained on-line for the entire month and operated at 100% power for the month of May 1997.
There were three load reductions which are identified below.
Power was reduced to 87% on May 11, 1997, starting at 0305 hours0.00353 days <br />0.0847 hours <br />5.042989e-4 weeks <br />1.160525e-4 months <br /> to perform monthly turbine valve testing. The unit was returned to 100% power on May 11, 1997, at 0714 bours.
1 Power was reduced to 80% on May 27, 1997, starting at i
1135 hours0.0131 days <br />0.315 hours <br />0.00188 weeks <br />4.318675e-4 months <br /> due to "C" Feedwater isolation resulting from a UPS failure. The unit was returned to 100% power on May 2
28, 1997, at 0229 hours0.00265 days <br />0.0636 hours <br />3.786376e-4 weeks <br />8.71345e-5 months <br />.
Power was reduced to 75% on May 31, 1997, starting at 0659 for scram time testing.
At the end of the month, the unit was at 76% power.
At the end of the month the unit had been on-line for 205 days.
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l DOCKET No.: 50-354
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UNIT: Hope Creek l
DATE: 06/05/97 I
COMPLETED BY:
?..
Kepley i
TELEPHONE: (609) 339-1106 l
SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS l
FOR THE HOPE CREEK GENERATING STATION i40 NTH:
MAY 1997 l
The following items completed during April 1997 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 cafety 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.
Design Changes Summary of Safety Evaluations Replacement of Air-Operated Valves in the Safety Auxiliary Cooling System (SACS), 4EC-03612 Pkg.
8, Revision O.
The Standby Diesel Generator (SDG) room cooler water supply valves in the SACS were replaced.
The previous valves were carbon steel air actuated flexible-wedge gate valves.
The replacement valves are stainless steel air actuated ball valves which are more suitable to the application.
UFSAR Figure 9.2-5 was changed to show the type and material of the replacement valves and actuators.
UFSAR Table 3.9-18 was changed to show the valve / actuator replacement type.
This replacement does not change the function of the valves or the function of the affected systems.
The change does not alter the ability of the SACS, the SDG Room Recirculation i
System, or the SDGs to perform their intended safety function.
The affected systems will function in accordance with the original design and licensing basis of the plant.
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l Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Replacement of Air-Operated Valves in the Safety Auxiliary Cooling System (SACS), 4EC-03612 Pkg. 17, Revision O.
The High Pressure Coolant Injection (HPCI) pump room cooler water supply valves in the SACS were replaced.
The previous valves were carbon steel air actuated flexible-wedge gate valves.
The replacement valves are stainless steel air actuated ball valves which are more suitable to the application.
UFSAR Figure 9.2-4 was changed to show the type and material of the replacement i
valves and actuators.
UFSAR Table 3.9-18 was changed to show the valve / actuator replacement type.
This replacement doen not change the function of the valves or the function of the affected 1
systems.
The change does not alter the ability of the SACS, the Equipment Area Cooling System (EACS), or the HPCI pump room unit coolers to perform their intended safety function.
The affected systems will function in accordance with the original design and licensing basis of the plant.
Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Replacement of Air-Operated Valves in the Safety Auxiliaries Cooling System (SACS), 4EC-03612 Pkgs. 20 and 23, Revision 0.
The Filtration, Recirculation, Ventilation System (FRVS)
Recirculation System cooling coil water supply valves in the SACS were replaced.
The previous valves were carbon steel air actuated flexible-wedge gate valves.
The replacement valves are stainless steel air actuated ball valves which are more suitable to the application.
UFSAR Figure 9.2-4 was changed to show the type and material of the replacement valves and actuators.
UFSAR Table 3.9-18 was changed to show the valve / actuator replacement type.
This replacement does not change the function of the valves or the function of the affected systems.
The change does not alter the ability of the SACS or the FRVS Recirculation System to perform their intended safety functions.
The affected systems will function in accordance with the original design and licensing basis of the plant.
Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Replacement of Air-Operated Valves in the Safety Auxiliaries l
Cooling System (SACS), 4EC-03612, Pkg. 30, Revision O.
The Core l
Ppray (CS) pump room cooler water supply valves in the SACS were replaced.
The previous valves were carbon steel air actuated flexible-wedge gate valves.
The replacement valves are stainless
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.o steel air actuated ball valves which are more suitable to the application.
UFSAR Figure 9.2-4 was changed to show the type and material of the replacement valves and actuators.
UFSAR Table j
3.9-18 was changed to show the valve / actuator replacement type.
j This replacement does not change the function of the valves or i
the function of the affected systems.
The change does not alter the ability of the SACS, the Equipment Area Cooling System
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(EACS), or the CS pump room unit coolers to perform their intended safety functions.
The affected systems will function in l
accordance with the original design and licensing basis of the plant.
Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Remove Flow Switch 1EPFS-2225A, 4HE-00356, Pkg, 1, Revision O.
This design change removed the service water screen and backwash flow switch, 1EPFS-2225A, and installed gaskets and a blind flange in place of the flow switch.
Stress and seismic evaluations have been performed to demonstrate the adequacy of the change.
The replacement of the flow switch with a blind flange does not affect the amount of flow being provided to the Reactor Auxiliary Cooling System (RACS) and Safety Auxiliary Cooling System (SACS) heat exchangers during plant conditions, i
transients, and accident conditions.
The change does not:
1) change, degrade, or prevent actions described or assumed in any accident described in the UFSAR, 2) alter any assumptions previously made in evaluating radiological consequences of any accident described in the UFSAR, 3) affect the mitigation of the radiological consequences of any accident described in the UFSAR,
- 4) affect a fission product barrier, or 5) change the composition or inventory of radioactivity releases.
This change does not adversely affect the operability of the Station Service Water System (SSWS).
1 Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
j UFSAR Change Notices Summary of Safety Evaluations UFSAR CHANGE NOTICE 97-015, Nuclear Engineering Department
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Organizational Changes.
This Updated Final Safety Analyis Report I
(UFSAR) change reorganizes the functions within the Nuclear i
Engineering Department.
The description of the entire Nuclear Engineering Department function is now consolidated under the control of the Senior Vice President - Nuclear Engineering.
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provides flexibility in transferring responsibilities between direct reports to the Senior Vice President - Nuclear Engineering.
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1 Therefore, this UFSAR change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
UFSAR CHANGE NOTICE 97-013, Nuclear Business Unit (NBU)
Organization Changes.
This UFSAR change shifts, clarifies, and consolidates responsibilities for NBU organizations, revises the I
succession of authority and responsibility for Salem station operations, establishes new positions, and corrects editorial errors.
The changes provide better control over NBU activities, l
and do not involve any reduction in the qualifications or training requirements, technical support, management control, commitments, or the quality assurance program.
i Therefore, this UFSAR change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Temporary Mociifications Summary of Safety Evaluations Temporary Modification 97-003, Leak Repair of Steam Tunnel Main Steam Drain Line, 1-AB-028-DBD-2".
This Temporary Modification installs clamps and a hanger to contain a leak on line 1-AB-028-DBD-2", a main steam drain line, shown in UFSAR Figures 6.2-4 6 l
and 6.3-1.
Line 1-AB-028-DBD-2" is non-safety related and
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provides a flowpath to the condenser in the standby mode of High Pressure Coolant Injection (HPCI)/ Reactor Core Isolation Cooling (RCIC). In the event that HPCI or RCIC is placed into service, the flow path of the main steam drain line is isolated.
This Temporary Modification does not alter or change the function or design requirements of the main steam drain line as described in the UFSAR. The clamp is designed to meet the temperature and pressure requirements of the main steam drain line.
A temporary j
hanger is installed to support the additional load of the clamps.
The main steam drain line meets seismic II/1 requirements with the temporary hanger installed.
Therefore, implementation of this temporary modification does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
Temporary Modification 97-004, Aligning Valves to Allow Continuous Draining of Drain / Bucket Trap, 1KLDT-126.
This Temporary Modification establishes an off normal line-up for the Primary Containment Instrument Gas (PCIG) drain trap, 1KLDT-126, on the "B" compressor skid.
The off normal line-up consists of i
l closing the trap inlet valve, closing the trap equalizer valve l
and throttling the trap bypass valve open.
This is a change to l
the normal configuration of the facility as described in the l
USFAR.
This line-up allows the water generated from the intercooler condenser / separator to be eliminated without i
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excessive PCIG loss.
Normal operation of the traps blows gas into the equipment drain funnel in the cor.pr essor area while expelling water.
This Temporary Modificat :nn establishes a controlled small bleed from the separator.
The continuous small (2.5 SCFM maximum) blowdown is a small increase to the normal total flow and any change to radiological conditions will be detected by existing plant monitoring systems.
Water elimination is included in the original design basis function of the system.
No unmonitored release path is created by this temporary modification.
t Therefore, implementation of this temporary modification does not increase the probability or consequences of an accidenc previously described in the UFSAR and does not involve an i
Unreviewed Safety Question.
Deficiency Reports Summary of Safety Evaluations l
Procedures Summary of Safety Evaluations Other Summary of Safety Evaluation There were no changes in these categories implemented during April 1997.
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