ML20128J997

Monthly Operating Rept for Jan 1993 for Hope Creek Generation Station,Unit 1

ML20128J997
Person / Time
Site:	Hope Creek
Issue date:	01/31/1993
From:	Hagan J, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9302180035
Download: ML20128J997 (13)

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O PSEG Pubhc Service Electnc and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station February 11, 1993 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 January are being forwarded to you with the summary of changes, tests, and experiments that were implemented during January 1993 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, or J. J. Ha an General Manager -

Hope Creek Operations RAR:1d g p Attachments #

C Distribution t

18:106 The Eneroy People f,(ST1 l\

9 9302180035 930131 PDR ADOCK 05000354 , , , , w.,; , m R PDR

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ __--___a

't e INDEX NUMBER SECTION OF PAGES Average Daily Unit Power Level. . . . . . . . . . . 1 Operating Data Report . . . . . . . . . . . . . . . 2 Refueling Information . . . . . . . . . . . . . . . 1 Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 6

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h.. -AVERAGE DAILY-UNIT-POWER LEVEL I

DOCKET NO.. 50-354

! UNIT Hope Creek l DATE 2/11/93 ,

! -COMPLETED BY V. Zabialski TELEPHONE- f609) 339-3506 l

MONTH January 1993

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! DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY-POWER-LEVEL (NWe-Net) (NWa-Net).

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! 1. 1064 17. 1064 l- i

j. 2. 1072 18. 1063 t

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3. 1061 19. 1080 1

! 4. 1912- 20. 1064

! 5. 1056 21. 1062 1

6. 1066 22. 121Q >

l 7. 1064 23. 221 1

'; 8. 1911 24. 1060 f' '9. 1068 25. 1061 4 . .

i- 10. 1063 .26. r1Q11

11. 1076 27. 1072 '

1 3 - 12 . 1062 28, 1065 1

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13. 1063 29. 1921' t

-14 , 1071 30. 1058

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.i 15 . - laEZ 31. 1052

16. 1065 '

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OPERATING DATA REPORT DOCKET NO. 50-354 UNIT HoDe Creek DATE 2/11/9.1 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 OPERATING STATUS

1. Reporting Period January 1993 Gross Hours in Report Period lii

2. Currently Authorized Power Level (MWt) liga 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 744.0 744.0 44.999.6

6. Reactor reserve shutdown hours 222 2x2 222

7. Hours generator on line 744.0 744.0 44.248.9

8. Unit reserve shutdown hours Az2 Q22 Das

9. Gross thermal energy generated 2.440.787 2.440.787 140.654.005 (MWH) "

10. Gross electrical energy 825,410 825,410 46.573.464 generated (MWH)

11. Het electrical energy generated 790.922 790.922 44.493.30.1 (MWH)

12. Reactor service factor 100.0 100.0 gata

13. Reactor availability factor 100.0 100.0 83.9

14. Unit service factor 100.0 122xn 82.5 l

15. Unit availability factor 100.0- 12Q22 82.5

16. Unit capacity factor (using MDC) 103.1 103.1 Rath

17. Unit capacity factor 9P16 99.6 77.7 (Using Design MWe) 18.-Unit forced outage rate 22A 1 02 dal 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 j

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OPERATING DATA REPORT l UNIT SHUTDOWNS AND POWER REDUCTIONS j DOCKET NO. 50-354

. UNIT. Hoom Creek-l DATE 2/11/93

COMPLETED BY V. Zabialski TELEPHONE (609) 339-3506-1

MONTH J.gnuary 1993 i METHOD OF i SHUTTING j .

DOWN THE'-

l TYPE REACTOR OR i F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS 4

None I

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9 REFUELING INFORMATION DOCKET NO. 50-35.4.

UNIT Hooe Creek DATE 2/11/93 COMPLETED BY L Hollingpy_prj;h TELEPHONE (609) 339-1011 MONTH January 1993

1. Refueling information has changed from last month:

Yes No X

2. Scheduled date for next refueling: 3/5/94

3. Scheduled date for restart following refueling: 4/23/94

4. A. 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?

Yes No X If no, when is it scheduled? 2/18/94

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

6. Important licensing considerations associated with refueling:

- Highly likely that will use same or similar fresh fuel as current cycle: no new considerations.

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage (prior to refueling) laga C. In Spent Fuel Storage (after refueling) 1232 to 1264

8. Present licensed spent fuel storage capacity: 4991 Future spent fuel storage capacity: 4006

9. Date of last refueling that can be discharged 11/4/ 2010 to spent fuel pool assuming the present (EOC16) licensed capacity:

(does not allow for full-core offload)

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HOPE CREEK GENERATING STATION

, MONTHLY OPERATING

SUMMARY

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? JANUARY 1993 a

j Hope Creek entered the month of January at approximately 100%

i power. The unit operated throughout the month without i experiencing any shutdowns or reportable power-reductions. As of f January 31, the plant had been on-line-for 57 consecutive days.

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i y SUMh.ARY OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION

, JANUARY 1993 4

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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 sr.fety previously evaluated in.the safety analysis report nay be 4 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.

I The 10CFR50.591 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 j plant effluent releases and did not alter the existing

environmental impact. The 10CFR50.59 Safety Evaluations l determined that no.unreviewed safety or-environmental questions are involved.

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. i QQE Descriotion of Safety Evaluation  !

4EC-1010/06 This DCP installed six high pressure sodium flood i light fixtures in the Torus room at elevation 77. I It also installed four convenience receptacles on the wall above the catwalk directly below four of the light fixtures.

This DCP provides better lighting / receptacle availability and enhances the personnel nafe working environment. Torus lighting is non-aafety related and non-class lE. The new light fixtures, receptacles, cable, and junction boxes meet the applicable seismic and separation criteria.

Therefore, this DCP does not involve any Unreviewed safety Questions.

4EC-3293/01 This DCP added a handrail in the Reactor Building around the center line of the Torus. The handrail is being supported from existing structural framing steel.

The new handrail is not integral with any operating plant system and does not contribute towards malfunction of any equipment important to safety.

Additionally, this DCP meets the applicable seismic requirements. Therefore, this DCP does not involve any Unreviewed Safety Questions.

4EC-3320/01 These DCPs rewired limit switches on motor operated 4EC-3320/02 valves. INPO SOER 86-02 documented industry operating experience showing that when the open (red) light for the valve actuated by a limit switch is on the same rotor as the open torque bypass switch, the red light would go out at the same time as the bypass switch actuation. Because the bypass switch was set to be closed from a valve position of almost full closed to full closed, a valve position indication could show the valve full closed when the valve was only almost full closed.

This DCP rowired the limit switches so that the red l light is actuated from a different rotor than the open torque switch bypass. The rotor that operates the light is set for full closed. The red light will not go out when the bypass switch closes, but only when the valve is full closed.

The failure of the new limit switch is no more probable than the failure of the previous limit switch. The new limit switch is standard equipment supplied with the operator. It was installed in i

the equipment, this DCP used a s I spared a switch that was in use. pare switch and Therefore, this DCP does not involve any Unreviewed Safety Questions.

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DCE Description of Safety Evaluation

4EC-3359/01 This DCP added a platform and widened an 6xisting

' platform-in-the Reactor Building. The-platforms are classified as Seismic Category II/I and are attached to.the seismic Category I drywell concrete wall.

The platforms are not integral with any operating plant system.and do not contribute towards the malfunction of any equipment important to safety.

Therefore, this DCP does not involve any Unreviewed Safety Questions..

4EX-3183 This DCP changed the.setpoint on the flow controllers on the Radwaste Tank. Vent Filters. Due to inaccuracies in the flow measurement introduced by flow in the. sensing lines, actual flow was-greater than measured flow.- Changing the setpoint '

reduced actual flow to. design flow. This DCP is a test to verify that filter life will be extended by-operating at design flow. '

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The Radwaste Tank Vent Filter traina1are not safety related. They play no part in the prevention or-mitigation of an accident. This DCP did not alter the design intent-of the system. Therefore, this DCP does not involve any Unreviewed Safety Questions.

4HE-0006 This DCP replaced a portion of the blowdown'line from the bottom of a Steam Seal Evaporator "T" connection to downstream of-a 90* elbow. The original piping was Schedule 80 carbon steel, it ,

was replaced with Schedule-40 stainless steel. '

The Steam Seal system does not have.any safety l related function and does not compromise.any safety-j~ related system or component. Therefore,Ethis DCP does not involve any Unreviewed Safety Questions.

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IMB Descriotion of Safety Evaluation'93-003 This TMR installed two jumpers to simulate a high i differential water level signal across the 'A' j Service Water Travelling Screen, inhibiting low

speed operation. This TMR will remain in place

until corroded conduit that is-associated with a l ' level element is replaced.

While the simulated high differential water level

! signal is in place, the screen will operate'in the

! high speed mode. The-failure of one Station

! Service Water Loop will not prevent the Reactor i from being brought to a safe-shutdown. Therefore,

! this TMR does not involve any Unreviewed Safety

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i i Procedure Revision -Descriotion of Safety Evaluation

! NC.NA-AP.ZZ-0058(Q) This procedure revision changed-titles and

Rev. 1 responsibilities-for the Nuclear Department Corrective Action Program. The title Vice

, President - Nuclear Engineering was changed

, to General Manager - Engineering and Plant Betterment, and the responsibility for the

, procurement of nuclear fuel was transferred i to the Vice President - Nuclear Operations.

The title General Manager - Procurement and Material Control was changed to General Manager - Material Control. Additionally, responsibilities were added/ revised for the

. Vice President - Nuclear Operations, the s General Manager - Nuclear Operations

! Support, General Manger - Nuclear Services, and the General Manager - Nuclear Human Resources.

This procedure revision includes I administrative changes only and does not 3

affect any system operation. Therefore, 4

this procedure revision did not involve

. Unreviewed Safety Questions.

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