LR-N970321, Monthly Operating Rept for Apr 1997 for Hope Creek Generating Station

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Monthly Operating Rept for Apr 1997 for Hope Creek Generating Station
ML20141F844
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 04/30/1997
From: Bezilla M, Kepley L, Todd F
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N970321, NUDOCS 9705220066
Download: ML20141F844 (12)
v • d • e


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.. O PSEG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit NAY 151997 LR-N970321 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 April 1997 are being forwarded to you with the summary of changes, tests, and experiments that were implemented during March 1997 pursuant to the requirements of 10CFR50.59(b).

Refueling-information for Hope Creek Generating Station is also provided. There is one Technical Specification change required related to refueling. There are also Trchnical Specification changes that are required to resolve licensing basis issues prior to the end of Hope Creek Generating Station's next refueling outage, which is currently scheduled to end on November 5, 1997.

Sincerely yours, Mark B. Bezilla General Manader -

Hope Creek Operations RH:LK:DS  !

l Attachments C Distribution 9705220066 970430 PDR ADOCK 05000354 R

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INDEX NUMBER M8SECTION OF PAGES Operating Data Report................................ 2 Average Daily Unit Power Level....................... 1 Refueling Information................................ 1 1 1

Monthly Operating Summary............................ 1  !

Summary of Changes, Tests, and Experiments............ 5 l

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DOCKET No.: 50-354 UNIT: Hope Creek DATE: 05/08/97 COMPLETED BY: F. Todd TELEPHONE: (609) 339-1316

. OPERATING DATA REPORT OPERATING STATUS

1. Reporting Period April 1997 Gross Hours in Report Period 719
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 Cumulative Month Date ,

5. No. of hours reactor was 719.0 2879 76602.1 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 719.0 2879 75439.0
8. Unit reserve shutdown hours 0.0 0.0 0.0
9. Gross thermal energy 2241308 9245544 241113791 generated (MWH)
10. Gross electrical energy 756650 3136620 80030833 generated (MWH)
11. Net electrical energy 725381 3014976 76486695 generated (MWH)
12. Reactor service factor 100.0 100.0 84.3
13. Reactor availability factor 100.0 100.0 84.3
14. Unit service tactor 100.0 100.0 83.0
15. Unit availability factor 100.0 100.0 83.0
16. Unit capacity factor (using 97.9 101.6 81.7 MDC)
17. Unit capacity factor (using 94.6 98.1 78.9 Design MWe)
18. Unit forced outage rate 0.0 0.0 4.5
19. Shutdowns scheduled over next 6 months (type, date, &

duration):

Refueling Outage, September 6, 1997, 50 days l

l 20. If shutdown at end of report period, estimated date of start-up:

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DOCKET NO.: 50-354 l UNIT: Hope Creek l DATE: 05/08/97 COMPLETED BY: F. Todd TELEPHONE: (609) 339-1316 r

OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS (

MONTH APRIL 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 1 4/4/97- S 56 Hrs. A 1 Repair Steam 4/7/97 Tunnel Leak 1

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DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 05/08/97

( COMPLETED BY: F. Todd l TELEPHONE: (609) 339-1316 l

AVERAGE DAILY UNIT POWER LEVEL I

MONTH APRIL 1997 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 1066 17 1052 2 1054 18 1059

3. 1056 19 1060 1

4 '1038 20 1057 5 284 21 1059 i

6 501 22 1050 j 7 1038 23 1076 8 1061 24 1048 9 1061 25 1078 10 1065 26 1034 l i

11 1039 27 1055 12 1047 28 1048 13 928 29 1C55 l 14 1075 30 1070 15 1059 31 N/A 16 1063

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DOCKET NO.: 50-354 f UNIT: Hope Creek  ;

DATE: 05/08'/97 {

COMPLETED BY: L. Kepley TELEPHONE: (609) 339-1106 REFUELING INFORMATION.  ;

7 MONTH APRIL'1997 ,

1. -Refueling information has changed from last month: l Yes- X_ No _
2. Scheduled date for next refueling (RF07) : 9/6/97
3. . Scheduled date for restart following refueling- '

11/5/97 3 4A. Will Technical Specification changes or other' license amendments be required? 1 1

Yes X No -

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I B. Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee (SORC)?

Yes No X If no, when is it scheduled? To Be Determined for Cycle 8 COLR

5. Scheduled date(s) for submitting proposed licensing action:

A License Change Request will be submitted to the NRC on or before June 30, 1997.  ;

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In;portant licensing considerations associated with refueling:

The above required-LCR concerns the Technical Specification Safety Limit Minimum Critical Power Ratio for Cycle 8. j

7. Number of Fuel Assemblies: i

.A. Incore 764 ,

B. 'In Spent Fuel Storage 1472

8. Present licensed spent fuel storage capacity: 4006 -

Future spent fuel storage capacity: 4iTO6

9. ~Date of last refueling that can be discharged 5/3/2006 to'. spent fuel pool assuming the present licensed capacity: i (EOCl3)

(Does allow for full-core off-load)

(Assumes 244 bundle reloads every 18 months until then)

(Does not allow for smaller reloads due to improved fuel) l

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l E DOCKET No.: 50-354-UNIT: Hope Creek DATE: 05/08/97 l

COMPLETED BY: .F. Todd I TELEPHONE: -(609) 339-1316 ]

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MONTHLY OPERATING

SUMMARY

MONTH- ' APRIL 1997. l J

  • The Hope Creek Generating Station remained on-line for- '

theLentire month'and operated at 100% power for the month ,

of April 1997. There were two load reductions which are l identified below. l

  • Power was reduced to 25% on April.4, 1997, starting at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />'to repair steam tunnel leak. The unit was '

returned to 100%-power on April 7, 1997, at 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br />. j

  • Power was reduced to 79% on April 12, 1997, starting at i 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br /> to perform rod adjustment. The unit was returned to 100% power on April 13, 1997, at 2230 hours0.0258 days <br />0.619 hours <br />0.00369 weeks <br />8.48515e-4 months <br />. .
  • 'At the end of the month the unit had been on-line for 174 i days.

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DOCKET NO.: 50-354 UNIT: Hope Creek DAT; 05/08/97 COMPLETED BY: L. Kepley TELEPHONE: (609)339-1106

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTH April 1997 The following items completed during March 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 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 snowed 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 ,

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i l . , .. i Design Changes Summary of Safety Evaluations l

Replacement of Moisture Separators on Primary Containment Instrument Gas (PCIG) Compressor Skid 1B-S-934, 4EC-03274 Pkg. 2, Revision O. Four (4) moisture separators were replaced on the PCIG skids (two separators per skid). The new separators were fabricated to meet PSE&G's Design / Detail Specification for moisture separators in the PCIG system, but could not be purchased with the appropriate ASME Section III N-symbol affixed as originally supplied. The >

installation of a "non-N-stamped" component to replace an "N-stamped" component is a change to the facility as described in Sections 3.2.2 and 9.3.6.3 of the UFSAR. The change in size, material and internal configuration of the moisture' separators does not alter the design bases of the PCIG system, nor are'the design parameters or the description of the PCIG system altered. However, the PCIG skid is listed as a safety related system in the UFSAR. The compliance with the ASME "N" stamped components ensures design margin. The intent of the "N" stamp safety related designation will be satisfied by compliance with the requirements of NRC Generic Letter 89-09, "ASME Section III Component Replacement", and ASME design criteria in both i Section III and VIII. The new replacement separators are  ;

superior in design and fabrication to the original -

components. This replacement does not impact the design .

basis requirements for the Main Steam Isolation Valve i Sealing system or the 10CFR100 offsite dose. )

l Therefore, this design change does not increase the l probability or consequences of an accident previously j described in the UFSAR and does not involve an Unreviewed i Safety Question. I Replacement of Air-Operated Valves in the Safety Auxiliaries 1 Cooling System (SACS), 4EC-3612, Pkgs 11, 14, 15, and 16, j Revision O. The Residual Heat Removal (RHR) 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 l steel air actuated ball valves which are more suitable to i the application. UFSAR Figure 9.2-4 was changed to show the type and material of the replacement valves / 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, l the Equipment Area Cooling System (EACS), or the RHR pump

room unit coolers to perform their intended safety l

functions. 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.

l Replacement of Air-Operated Valves in the Safety Auxil2. aries  !

Cooling System (SACS), 4EC-3612, Pkgs 25, 28, and 31, Revision O. The Core Spray (CS) pump room cooler water i 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 / actuators. i 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 Equipment Area Cooling System (EACS), or the CS pump ,

room unit coolers co 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 i Safety Questj on.

Replacement of Air-Operated Valves in the Safety Auxiliaries i Cooling System (SACS), 4EC-03612 Pkgs. 19, and 21, Revision O. The Filtration, Recirculation, Ventilation System (FRVS)

Recirculation System cooling coil water supply valves in the i 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 l are-more suitable to the application. UFSAR' Figure 9.2-4 I was changed to show the type and material of the replacement valves / 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 l 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. ,

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I Temporary Modifications Summary of Safety Evaluations l

l Temporary Modification 96-023, Replacement of Safety Auxiliary Cooling System (SACS) valves 1EGV-545 and 1EGV-547

is to automatically close on a Loss of Coolant Accident '

l (LOCA) signal, Loss of Power (LOP) signal or a low-low-low  ;

SACS expansion tank level signal. The purpose of performing  ;

this temporary modification is to isolate the SACS cross-connect lines until the two motor operated valves in each .

line can be replaced. The SACS valves in the cross-connect  :

line are not leak tight when they are closed. If a break occurs in one SACS loop, these valves are designed to 1 automatically close on low-low-low SACS expansion tank level to protect the intact loop. Also, these valves are designed I to close on a LOCA or LOP signal. -This temporary  ;

modification will perform that safety function of the valves until the valves are replaced. The blank flanges in the ,

temporary flange assemblies are equivalent to the other SACS l piping in the Fuel Pool Cooling and Cleanup System (FPCCS)  ;

cross-connect ljnes. A piping stress analysis concluded that the pipe stresses were acceptable with the temporary flange assemblies installed. The temporary modification allows each FPCCS heat exchanger to be cooled by only one SACS loop. However, heat removal requirements are within the capacity of one FPCCS heat exchanger. The safety 1 function of the valves is assured by the temporary modification, l

Therefore, this temporary modification does not increase the I probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.

j Procedures Summary of Safety Evaluations NC . EP-EP . ZZ-0801 (Q) , Revision 0, Emergency News Center I Changes. The Nuclear Business Unit (NBU) Emergency Plan, l which is discussed in Section 13.3 of the Salem and Hope Creek Updated Final Safety Analysis Reports (UFSAR), has been changed to reduce the initial call-out staffing for the Emergency News Center (ENC) from 26 to 22 personnel. These changes do not result in a reduction in the positions listed .

on the UFSAR Emergency Plan Table 3-2, Nuclear Business Unit l Correlation to Supplement #1 of NUREG-0737, Table 2, and NUREG-0654, Table B-1. The ENC positions that have been

! l eliminated are not addressed in any regulatory requirement and do not affect plant safety.

r Therefore, implementation of this change does not increase the probability or consequences of an accident previously l described in the UFSAR and does not involve an Unreviewed Safety Question. r Other Summary of Safety Evaluation l

Safety Evaluation H97-014, Revision 1, Non-use of Drywell Cooling Post Loss of Coolant Accident (LOCA) . This change administratively prohibits the restoration of drywell cooling following a LOCA. Engineering Evaluation, H-1-GB- 1 MEE-1157, was performed to document the assessment of the  ;

post LOCA use of drywell coolers at Hope Creek. This evaluation determined that, if the containment isolation signals are overridden for the chilled water penetrations 4 after an accident, a water hammer event may occur and impact '

primary containment integrity. Since drywell cooling is J isolated on a LOCA signal, this change is consistent with ,

the Hope Creek design, in that the UFSAR does not credit the '

use of the drywell coolers for any event that results in an automatic isolation of the primary containment. UFSAR Section 6.2.2 indicates that the Residual Heat-Removal system provides the safety related means for removing heat from the primary containment during and after a LOCA.

However, interlock bypass features are provided in Emergency l Operating Procedures (EOPs) to allow cooling to be manually  ;

realigned to the drywell if desired. This change prohibits this realignment of drywell cooling, which differs from the direction provided in the BWR Emergency Procedure Guideline ,

approved by the NRC in 1988. The change does not affect the '

probability of a LOCA or any other event evaluated in the SAR. This change is consistent with the analysis in Chapters 6 and 15 of the UFSAR. Also, no other analysis )

credits the use of drywell cooling following a LOCA. l Therefore, implementation of this 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 Notices Summary of Safety Evaluations Deficiency Reports Summary of Safety Evaluations There were no changes in these categories implemented during l March 1997. l

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