ML20115E849

Monthly Operating Rept for June 1996 for Hope Creek Generation Station

ML20115E849
Person / Time
Site:	Hope Creek
Issue date:	06/30/1996
From:	Harris R, Lyons D Public Service Enterprise Group
To:
Shared Package
ML20115E848	List: ML20115E844 ML20115E849
References
NUDOCS 9607160257
Download: ML20115E849 (9)

Text

DOCKET NO.: 50-354 UNIT: Hooc Creek DATE: 7/02/96 COMPLETED BY: R. Harris TELEPHONE: W)9) 339-1777 OPERATING DATA REPORT OPERATING STATUS

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

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) Hg_ne n

4. Reasons for restriction (if any)

This Month yr To Date Cumulative S. No. of hours reactor was critical 720.0 2505.2_ 69429.1

6. - Reactor reserve shutdown hours M M M

7. Hours generator on line 720.0 2347,4 68289.0

8. Unit reserve shutdown hours 0.0 0.0 0.0

9. Gross thermal energy generated (MWH) 2211700 7263028 218037277

10. Gross electrical energy generated (MWH) 732990 2431671 72257293

11. Net electrical energy generated (MWH) 701737 2303599 69020834

12. Reactor service factor 100.0 17 4 83.1

13. Reactor availability factor 100.0 57.4 83.1

14. Unit service factor 100.0 53.8 81.7

15. Unit availability factor 100.0 53.8 81.7

16. Unit capacity factor (using MDC) 94.5 iL2 80.1

17. Unit capacity factor (using Design MWe) 91.3 49.4 77.4

18. Unit forced outage rate M M M

19. Shutdowns scheduled over next 6 months (type, date, & duration):

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

9607160257 960711 PDR ADOCK 0S000354 R PDR

DOCKET NO.: 50 354  ;

UNIT: Hope Creek DATE: 7/02/96 COMPLETED BY: R. Harris TELEPHONE: L609) 339-1777  ;

OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS MONTH JUNE 1996 METHOD OF SHUTTING DOWN THE

- i TYPE REACTOR OR 1 F= FORCED DURATION REASON REDUCING CORRECTIVE l NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS l

1 6/7/96 S 0 A-EHC 5 - Power Unit was reduced from

, thru Power Reduction 100% power to replace 6/9/96 Supply >20% EHC Power Supply, Replace-ment 2 6/22/96 S 0 H - Deep 5 - Power Unit was reduced from thru Shallow Reduction 100% power to perform j 6/26/96 Rod Swap >20% Deep Shallow Rod Swap

& Turbine & Monthly Turbine Tests Tests. Theload reduction was extended when "C" TIP failure occurred.

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9 DOCKET NO.: 50-354 UNIT: Lione creek l DATE: 7/02/96 COMPLETED BY: R. Ilarris TELEPHONE: (609) 339-1777 AVERAGE DAILY UNIT POWER LEVEL

! MONTH JUNE 1996 l

l DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1 1050 17 1025 2 1051 18 1038 3 1049 19 1034 l

4 1045 20 1029 5 1040 21 1031 l

6 1045 22 689 7 1015 23 785 8 237 24 834 I 9 909 25 895 ,

10 1040 26 1030 1

11 1037 27 1023 -

12 1030 28 1060 ,

i 13 1031 29 1044 ]

14 1035 30 1041 15 1034 31 N/A 16 1034 r

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l DOCKET NO.: 50-354 l UNIT: Hope Creek l DATE: 7/02/96 COMPLETED BY: R. Harris TELEPHONE: (609)339-1777 l

REFUELING INFORMATION MONTH JUNE 1996

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)?

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:

l Ho1 required.

6. Important licensing considerations associated with refueling:

N/A

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: 4006

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

(Does allow for full-core off-load)

(Assumes 244 bundle reloads every 18 months until then)

(Does no.1 allow for smaller reloads due to improved fuel) l l

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1 DOCKET NO.: 50-354 UNIT: Hooe Creek DATE: 7/02/96 COMPLETED BY: R. Harris .

TELEPHONE: (609) 339-1777 MONTHLV OPERATING

SUMMARY

l MONTH JUNE 1996 i e The Hope Creek Generating Station remained on-line for the entire month and operated at essentially 100% power for the month of June 1996 except two load reductions identified below.

• Power was reduced to 20% on June 7,1996 for EHC Power Supply Replacement.

The unit was returned to 100% power on June 9,1996.

• Power was reduced to 60% on June 22,1996 to perform Deep Shallow Rod Swap and performance of monthly turbine valve testing. The load reduction was extended when the "C" TIP failure occurred. The unit was returned to 100% power on June 26, l 1996. '

. At the end of the month the unit had been on-line for 98 days.

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DOCKET NO.: 50 354 UNIT: Hope Creek DATE: 06/26/96 COMPLETED BY: D. W. Lyons TELEPHONE: (609) 339-3517

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTil JUNE 1996  ;

The following items completed during May 1996 have been evaluated to determme.

1. If the probability of occurrence or the consequences of an accident or malfunction of j 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 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 efIluent 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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Design Changes Summary of Safety Evaluations e 4EC-3546: PACKAGES 13 & 14, STATION SERVICE WATER SYSTEM STRAINER VENT I,INE PIPING REPI,ACEMENTS This design change replaced one inch carbon steel Station Service Water System (SSWS) strainer vent lines in the Service Water Intake Structure with molybdenum-austenitic stainless steel alloy piping.

The replacement piping is designed and installed to the same design and licensing criteria as the original pipmg. UFSAR Figure 9.2-2, Sheet I will be modified to reflect the change of material. The replacement will not affect the Safety Class 3, Seismic Category 1 designation of the piping vent lines. Hydraulically and functionally the vent flows are the same and the SSWS flow delivered to the RACS and SACS Heat Exchangers are unaffected. The material replacement does not introduce a ? ipe break concern and postulated failures of these lines cannot adversely affect safety-re'ated stmetures, systems, l or components.

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.

e 4IIE-0346: PACKAGES 1 & 2, MODIFICATION TO SSWS BEEHIVE GRATES This design change provides more open flow area to the Station Service Water System (SSWS) emergency bypass lines by extending outward the non-safety related ' beehive grates." This change reduces the flow coeflicient of the ' beehive grate"which facilitates the SSWS in providing cooling water to the SACS and RACS Heat Exchangers under accident conditions following failure of the non-seismic SSWS discharge line to the Cooling Tower. This change was implemented as part of the corrective actions when a non-conservative calculational error in Ultimate Heat Sink temperature was identified.(AR 960403215). Extending the grates will not adversely impact the safety-related pip,ing to security barrier to prevent unauthorized entry through a vital barrier. Failures m the SSWS and other interfacing systems are bounded by previous analyses. The increase in SSWS flo v through the Emergency Bypass Line is within the design basis.

Therefom, 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.

UFSAR Change Notices Summary o.f Safety Evaluations e UFSAR CHANGE NOTICE UFSAR SECTION 17.2, OUALITY ASSURANCE DURING Tile OPERATIONS PilASE This change moves existing responsibilities between managers within the Quality Assurance / Nuclear Safety Review organization.

Specifically, Quality Verification will be the responsibility of the Manager - Corrective Action / Quality Services and QA - Programs will be assigned to the Manager - Quality Assessment. These changes will allow the department to operate more efliciently. These changes are administrative in nature and have no effect on failure modes, operational transients, postulated design basis accidents, or assumptions contained in the plant safety analysis.

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 CN 96-34, MAIN TURHINE CIV TESTING FREOUENCY FROM WEEKLY TO MONTHLY This change notice extends the frequency for testing of the main turbine Combined Intermediate Valves (CIV) stated in UFSAR Section 10.23.6 to monthly from weekly. The method of testing is not changed.

This change is based on recommendations from General Electric (GE). GE's recommendations came from a joint BWR Owners' Group and GE study, NEDO-32155,

' Technical Bases for Extending GE Turbine Valve Surveillance"and a study done for PSE&G to evaluate other testing frequencies to minimize load reductions. No plant parameters are affected by this change. Missile probabilities for the turbine remain within the criteria of Regulatory Guide 1.115 and NUREG 1048. There are no affects on equipment important to safety. The turbine-generator has no safety-related function.

Failure of the turbine-generator does not compromise any safety-related system or component or prevent the safe shutdown of the plant.

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.

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4 4 Other Summary of Safety Evaluation e

ENGINEERING EVALUATION H-1-VAR-MEE-1997. DCP EXCLUSION ZONE FOR CIRC WATER / SWITCHGEAR BUILDINGS & ASSOCIATED HVAC SYSTEMS The purpose of this safety evaluation is to identify systems, structures and i components (SSC) of the Circulating Water / Switchgear Buildings and Associated HVAC l

1 Systems as being outside of the scope of nuclear jurisdiction and justify creation of a design change exclusion zone for these SSCs. The Design Change Process (DCP) will no longer a,pply to modifications to these SSCs.. These items will be deleted from UFSAR

descriptions and figure 9.4-18 (P&ID M-95-0). The systems affected are

Circulating i Water, Circulating Water Pumphouse and Switchgear Buildings, Associated HVAC l Systems, and Non IE 480 Volt MCC AC Power. These systems are not safety related l and not required for safe shutdown of the plant. Failure of these systems, as described in l UFSAR Sections 9.4.8 3 and 10.4.53, will not compromise any safety related system or component or prevent a safe shutdown of the reactor. The reclassification of these  ;

systems does not change any design or operating parameters. No work will be done  !

within the exclusion zone that will have impact beyond the zone boundary unless it is '

under control of the Design Change Process.

! Therefore, implementation of this Engineering Evaluation and corresponding UFSAR I change do 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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1 e UFSAR SECTION 13.6. SALEM-HOPE CREEK SECURITY PLAN. REVISION 7 l

l' The change implements Amendment No. 95 to the Hope Creek Technical Specifications.

This amendment deletes the requirements for SORC review of the Security Plan and implementing procedures and an annual audit of the Security Program from the Technical Specifications provided the requirements are added to the Security Plan. The SORC review requirement is added to the introduction of the Security Plan. The phrase 'bnder the cognizance of the Off-Site Safety Review stafl"was added to the audit requirement in section 13.1 of the Security Plan. No plant parameters or systems are affected by this change. There are no failure modes associated with this change. No plant operational transients or postulated design basis accidents are applicable to this change. Because the Security Program is designed to prevent purposeful acts of radiological sabotage, there is no Security accident analysis. This change is administrative only..

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

Temporary Modifications Summary of Safety Evaluations i

- Deficiency Reports Summary of Safety Evaluations Procedures Summary of Safety Evaluations There were no changes in these categories implemented during May 1996. -l l

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