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Monthly Operating Rept for Nov 1996 for Hope Creek Generation Station Unit 1
	ML20133B790
Person / Time
Site:	Hope Creek
Issue date:	11/30/1996
From:	Bezilla M, Harris R, Kepley L; Public Service Enterprise Group
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	LR-N96410, NUDOCS 9701060180
	Download: ML20133B790 (12)
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4 _ , a . .t __L s a 4 O PSEG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 0236 Nuclear Business Unit DEC 131996 LR-N96410 1

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U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 i

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 1996 are being forwarded to you with the summary of changes, tests, and experiments that were implemented during October 1996 pursuant to the requirements of 10CFR50.59(b).

4 Sincerely yours,

,j'Y b <Y Mark B. Bezilla General Manager -

Hope Creek Operations

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i NUMBER SECTION OF PAGES i i

Operating Data Report. . . .. .. . . . . . . . . . . . ..2 .

l. Average Daily Unit Power Level . . .. . . . . . ... .. . . .1. :

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! Refueling Information.... , . . . . . . . . . . .. . . . . . . . . . .1 ,

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

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

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i DOCKET NO.: 50-354 l UNIT: Hooc Creek DATE:- 12/10/96 l l COMPLETED BY: R. Harris l TELEPHONE: (609) 339-1777 i

l OPERATING DATA REPORT OPERATING STATUS

1. Reporting Period November 1996 Gross Hours in Report Period 720

2. Currently Authorized Power Level (MWt) 31 91 Max. Depend. Capacity (MWe-Net) 1031 Design Electrical Rating (MWe-Net) 10_6.1

3. Power Level to which restricted (if any) (MWe-Net) None

4. Reasons for restriction (if any)

This Month Yr To Date Cumulative

5. No. of hours reactor was critical 597.0 6055.2 72979.1

6. Reactor reserve shutdown hours 0.0 0.0 0.0

7. Hours generator on line 574.0 5874.4 71816.0

8. Unit reserve shutdown hours OJ 00 01 0

9. Gross thermal energy generated (MWH) 1824741 18675222 229449471

10. Gross electrical energy generated (MWH) 618290 6246111 76071733

11. Net electrical energy generated (MWH) 590776 5963841 72681076 ,

12. Reactor service factor 82.9 75.3 83.7

13. Reactor availability factor 82.9 213 83.7

14. Unit service factor 79.7 73.1 82.3

15. Unit availability factor 79.7 73.1 82.3

16. Unit capacity factor (using MDC) 79.6 71.9 80.8

17. Unit capacity factor (using Design MWe) 76.9 69.5 78.1

18. Unit forced outage rate 0.0 0.0 4.7 l 19. Shutdowns scheduled over next 6 months (type, date, & duration):

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

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DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 12/10/96 COMPLETED BY: R. Harris TELEPHONE: (609) 339-1777 OPERATING DATA REPORT UNIT SIIUTDOWNS AND POWER REDUCTIONS MONTII NOVEMBER 1996 METHOD OF SHUTTING j DOWN THE l TYPE REACTOR OR I F= FORCED DURATION REASON REDUCING CORRECTIVE l NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS

1. I1/1/96 S 146 hours0.00169 days 0.0406 hours 2.414021e-4 weeks 5.5553e-5 months B-Plan.t ' l-Manual The reason for this to Maintenance planned maintenance was 11/8/96 to repair the reactor recirc pump seal. The seal was repaired 2 days ahead of schedule. Shedown commenced @ 1803 hours0.0209 days 0.501 hours 0.00298 weeks 6.860415e-4 months on 11/1/96. Generator taken off-line @ 0441 hours0.0051 days 0.123 hours 7.291667e-4 weeks 1.678005e-4 months on 11/2/96. Power ascension commenced (Rx critical) at 0225 hours0.0026 days 0.0625 hours 3.720238e-4 weeks 8.56125e-5 months on 11/6/96, synch'd to the grid at 0117 hours0.00135 days 0.0325 hours 1.934524e-4 weeks 4.45185e-5 months on 11/8/96. Achieved 100%

power at 1521 hours0.0176 days 0.423 hours 0.00251 weeks 5.787405e-4 months on 11/9/96.

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DOCKET NO.: 50-354 UNIT: Hooc Creek DATE: 12/10/96 COMPLETED BY: R. Harris TELEPHONE: (609) 339-1777 AVERAGE DAILY UNIT POWER LEVEL I MONTH NOVEMBER 1996 i

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) l 1 937 17 1068 2 ;2 18 1062 3 Q 19 1061 4 Q 20 1062 5 9 21 1068 6 Q 22 1063 7 0 23 1067 8 535 24 1062 9 975 25 1075 10 1041 26 1047 11 1065 27 1062 12 1066 . 28 1069 13 1063 29 qq61 14 1076 30 1059 15 1038 31 N/A >

16 1091 .

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DOCKET NO.: 50-354 ,

UNIT: llope Creek DATE: 12/10/96 COMPLETED BY: L. Keolev TELEPHONE: (609) 339-1106 REFUELING INFORMATION MONTH NOVEMBER 1996 i

1. Refueling information has changed from last month:

Yes __ No X

2. Scheduled date for next refueling (RF07): 9/6/97

3. Scheduled date for restart following refueling: 11/5/97 4A. 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 (SORC)? 1 1

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

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

N_et required.

6. Important licensing considerations associated with refueling: ,

N/A

7. Number of Fuel Assemblies:

A. Incore 764 I B. In Spent Fuel Storage 1472 !

8. Present licensed spent fuel storage capacity: 4006 ;

Future spent fuel storage capacity: 4006 j

9. Date oflast refueling that can be dischary,ed 5/3/2006 to spent fuel pool assuming the present hcensed capacity: (EOCl3)

(Does allow for full-core off-load)

(Assumes 244 bundle reloads every 18 months until then)

(Does nat allow for smaller reloads due to improved fuel)

DOCKET NO.: 50-354 UNIT: HODC CTCCk DATE: 12/10/96

, COMPLETED BY: R. Harris i TELEPHONE: (609)339-1777 i

MONTHLY OPERATING

SUMMARY

MONTH NOVEMBER 1996

The Hope Creek Generating Station remained on-line at 100% power until November 1,1996 when it began a scheduled one week planned maintenance outape. The reason i for this )lanned maintenance outage was to repair the reactor recirculation pump seal.

, The sea: was repaired 2 days ahead of schedule. Shutdown commenced @ 1803 hours0.0209 days 0.501 hours 0.00298 weeks 6.860415e-4 months

on 11/1/96. The generator was taken off-line @ 0441 hours0.0051 days 0.123 hours 7.291667e-4 weeks 1.678005e-4 months on 11/2/96. Power ascension commenced (Reactor critical) at 0225 hours0.0026 days 0.0625 hours 3.720238e-4 weeks 8.56125e-5 months on 11/6/96, the unit was synchronized to the grid at 0117 hours0.00135 days 0.0325 hours 1.934524e-4 weeks 4.45185e-5 months on 11/8/96. 100% power was achieved at

1521 hours0.0176 days 0.423 hours 0.00251 weeks 5.787405e-4 months on 11/9/96.

j e The unit remained on-line for the entire month after 11/9/96 and operated at essentially 100% power except two load reductions identified below.

j e Power was reduced to 90% on November 10.1996 starting at 1703 hours0.0197 days 0.473 hours 0.00282 weeks 6.479915e-4 months to perform j SCRAM time testing on control rod 42 07. The unit was returned to 100% power on November 10,1996 at 2208 hours0.0256 days 0.613 hours 0.00365 weeks 8.40144e-4 months .

Power was reduced to 97% on November 15,1996 starting at 0802 hours0.00928 days 0.223 hours 0.00133 weeks 3.05161e-4 months to perform control rod pattern adjustment. The unit was returned to 100% power on November 15,1996 at 0839 hours0.00971 days 0.233 hours 0.00139 weeks 3.192395e-4 months .

. At the end of the month the unit had been on-line for 23 days.

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DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 12/10/96 COMPLETED BY: L. Keolev TELEPHONE: (609) 339-3517

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR TIIE IIOPE CREEK GENERATING STATION MONTH NOVEMBER 1996 1

The following items completed during October 1996 have been evaluated to determine:

1. If the probability of occurrence or the consequences of an accident or malfunction of l 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 diiTerent 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 eflluent releases and did not alter the existing environmental impact. The 10CFR$0.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.

Design Changes Summary of Safety Evaluations 4EC-03629, Package 1, Process Radiation Monitoring - South Plant Vent Radiation Monitering System Flow Measurement Loop Changeout. This Design Change Package (DCP) replaced the existing South Plant Vent (SPV) pneumatic flow l measurement circuit with an electronic flow measurement instrument. The SPV flow l measurement is required by Technical Specifications, however, the SPV Radiation Monitoring System (RMS) is a non-safety related system. Since the system is non-safety ,

related, no credit is taken for it in the accident analyses in the UFSAR. There are no i interfaces with safety related structures, systems, or components (SSCs).

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 i Question.

4EC-03296, Package 1, Fire Detection Computer Replacement Project. This DCP replaced the Pyrotronics Multilarm V Computer System located in the control room and the Pyrotronics Transmitters / Receivers with Foxboro Intelligent / Automation (I/A) Series equipment, a distributed control system (DCS). The new equipment being installed by this modification is being used in an application for which it is designed and has been situated relative to other equipment such that it does not increase the likelihood of a fire. In addition, the likelihood that a fire will occur is also related to factors such as administrative controls that are in place for use and handling of combustible materials and use ofignition sources. This change does not impact these administrative controls. The signaling line circuits are run in conduit (just as the original design) or are metal jacketed which means the probability of a wiring failure and the probability of a fire coincident with a cable break remains the same. System reliability has been increased with the use of the

internal system diagnostics and the fault tolerant design of the Foxboro I/A Series equipment. The fire alarm system provides monitoring and alarm functions only and does not provide control function such as actuation of a fire suppression system. The system has no interface with safety related or important to safety systems. This change provides adequate protection for the affected equipment. Loss of the Fire Detection Computer will not prevent the actuation of a fixed fire protection system. The equipment being installed per this modification meets the environmental condition for the affected areas, and is installed in accordance with seismic H/l requirements were applicable and properly supported such that physical limitations of existing supports are not exceeded. This ;

modification does not increase the radiological consequences of any of the accidents '

evaluated in the UFSAR because the change does not: 1) alter any assumptions l previously made in the evaluation of the radiological consequences of any accident described in the UFSAR, 2) affect the fission product barrier, or 3) restrict access to vital areas or otherwise impede actions to mitigate the consequences of a reactor accident.

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.

4HM-0258-2, Permanent Lifting Points. This change added a new Overhead Heavy Load Handling System (OHLHS). This design change provides permanent lifting points attached to overhead structural steel to allow removal of a valve and its motor operator.

The designated weight of a heavy load for Hope Creek is 1,200 pounds or more. The valve and its motor operator each weight more than 1,200 pounds and are considered heavy loads. This change did not involve work on safety-related equipment or items interfacing with safety related equipment. Design reliability is demonstrated by analysis and material test reports or certificates of compliance. Proper fabrication and assembly is demonstrated by NDE of major load bearing welds and QC or field engineering inspections. The safety load path for the valve and its operator does not pass over any safety-related or safe shutdown equipment. In the event of a heavy load drop, no safety- l related or safe shutdown equipment will be affected. The design change is in compliance with NUREG-0612, Section 5.1, Evaluation Criteria IV.

Therefore, this design change does not increase the probability or consequences of an i accident previously described in the UFSAR and does not involve an Unreviewed Safety l Question. j

, UFSAR Change Notices Summary of Safety Evaluations t

UFSAR Change Notice # 96-93, Quality Assessment (QA) Reporting Relationship to Director Quality Assessment / Nuclear Safety Review (QA/NSR). The re?orting relationship of the Quality Assessment function to the Director-QA/NSR is being caanged to improve direction and control, oversight, and accountability. Currently, there are two separate Quality Assessment organizations responsible for implementation of the independent assessment program at the Hope Creek and Salam Generating Stations, respectively. The change consists of combining the two organizations into one under the Manager - Quality Assessment. The single organization will retain all of the responsibilities of the two original organizations. This change in the reporting relationship of the Quality Assessment function to the Director - QA/NSR has no direct effect on any systems, structures, components described in the Hope Creek and Salem UFSARs. The change does not reduce the level of personnel training and qualifications, management 4

control and authority, or QA Program commitments. The QA Program will continue to satisfy the criteria of 10CFR50, Appendix B upon incorporation of this change.

i Therefore, this UFSAR change does not increase the probability or consequences of an ,

accident previously described m the UFSAR and does not involve an Unreviewed Safety Question.

I UFSAR Change Notice #96-106, Relocation of Equipment Root Cause Analysis j Program Sponsorship Within Nuclear Engineering. The UFSAR change reflects a reorganization within the Nuclear Engineering Department. Hope Creek UFSAR Section 13.1.1.2.1.3 will no longer describe that the equipment root cause analysis is performed within Specialty Engineering. The reference is being deleted. The program sponsorship will reside within Engineering Assurance in Design Engineering and Projects but will not be described in the UFSAR. Root cause analysis of equipment failures will continue to be performed in accordance with the requirements of the Nuclear Business Unit's Corrective l Actiou Program. This change is admmistrative in nature and has no impact on the design or licensing basis. The technical and programmatic requirements surrounding root cause analysis of equipment failures have not changed.

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. )

i Temporary Modifications Summary of Safety Evaluations Temporary Modification 96-025, Battery IBD411 Cell No.19 Jumpered Out. This temporary modification addresses the jumpering out of battery cell number 19 in 125 VDC Class lE battery IBD411 and allows continued operability of the battery with 59 cells. Cell number 19 was disconnected from the battery bank and a cable jumper was installed between cells 18 and 20. The battery is designed to provide power to safety related equipment to mitigate the consequences of an accident. The change affects the battery voltage available to safety related loads in the 'B' channel, however, the design basis voltage to safety related loads will not be compromised. The remaining 59 cells will be sufficient to meet the minimum 108 V battery terminal voltage criteria to the Class IE loads.

Therefore, 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 96-028, Temporary Modification for Ventilation Ducting for Chemistry Hot Laboratory Fume Exhaust Hood for Inductively Coupled Plasma (ICP). This temporary modification installs flexible metallic ducting between the chemistry laboratory exhaust ventilation hood and the ICP chemistry analysis machine. The ICP machine is used for chemistry analysis associated with fuel warranty measurements. The temporary modification will not create an adverse impact to the chemistry laboratory exhaust ventilation system. The chemistry laboratory exhaust ventilation system is a non-safety related system and the installation of the flexible ducting will not impact safety-related systems. The function of this system is to maintain air flow from areas oflower radioactivity to areas of greater radioactivity. This temporary modification is in the

chemistry laboratory area only and does not effect components or equipment within the

)lant. The chemistry laboratory exhaust system is part of the auxiliary building radwaste 1 eating, ventilation and air condition (HVAC) system which is non-safety related. Any normal or abnormal source contribution from the chemical laboratory exhaust system will be small, with releases within the limits specified in 10CFR20 and the As Low As Reasonably Achievable (ALARA) guidelines of Appendix I of 10CFR50. Failure of the chemistry laboratory exhaust system does not compromise safety-related systems / components or prevent a safe shutdown of the plant.

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

Procedures Summary of Safety Evaluations NC.NA-AP.ZZ-0003(Q), Document Management Program. The procedure change replaces Senior Vice President - Nuclear Engineering with the Director - Nuclear Business Support as responsible for document control and records management activities. The change reflects organization structure and responsibility changes. The changes have no direct effect on any systems, structures or components described in the UFSAR. The changes are administrative in nature and have no direct effect on equipment failure modes.

The changes do not result in reduction in the qualifications and training requirements of personnel, the management control and authority, administrative program commitments or quality assurance program. The changes meet the intent of the Standard Review Plan, NUREG -0800, and other NRC requirements previously committed to in the UFSAR.

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

Other Summary of Safety Evaluation Security Plan, Revision 8, Hope Creek UFSAR Section 13.6. The Security Plan is designed to prevent purposeful acts of radiological sabotage. The changes to the Security Plan involve: 1) secunty organizational changes, 2) clarification oflanguage describing the basis of the vital area configuration of Salem Generating Station to establish that the configuration is the all-inclusive and licensed configuration, 3) Salem control room access portals, 4) additional language added relative to the land vehicle barrier system, 5) the addition of a provision for the electronic submittal of area access authorizations, and 6) deletion ofprovisions for temporary designated vehicles. The organizational change has no effect upon the performance of the functions of the medical and site access groups as they related to required physical exams and the Fitness-For-Duty program or required training.

The elimination of the temporary designated licensee vehicles (TDLV), in accordance with the molicy promulgated in NRC document 'SECY-93-326,

Subject:

ReconsideraHon of Nuc ear Power Plant Security Recuirements Associated with an Internal Threat", is a program enhancement. The only plant equipment affected by the changes are the Salem control room security doors. Malfunctions of door alarms and door hardware problems are promptly recognized and address by active security patrols. These changes are made in accordance with the provisions of 10CFR50.54(p), and do not reduce the safeguards effectiveness of the Security Plan. The changes were examined with regard to the Design Basis Srcurity Threat described in 10CFR73.l(a). The changes do not decrease the level of

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protection against the design basis security threat and therefore, the change will not ;

contribute to increased vulnerability to successful threat execution. !

' Therefore, this Security Plan revision does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewe<.' Safety Question. ,

Deficiency Reports Summary of Safety Evaluations i

There were no changes in this category implemented during October 1996.

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ML20133B790

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