ML20059G394

Monthly Operating Rept for Dec 1993 for Hope Creek Generation Station Unit 1

ML20059G394
Person / Time
Site:	Hope Creek
Issue date:	12/31/1993
From:	Hollingsworth, Hovey R, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9401240140
Download: ML20059G394 (12)

Text

. . . . _ _ _

i O PSEG i Pubhc Service Electric and Gas Cornpany P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station January 14, 1994 U. S. Nuclear Regulatory Commission Document Control Desk

• Washington, DC .20555 MONTHLY OPERATING REPORT- f HOPE CREEK GENERATION STATION UNIT 1 DOCKET NO. 50-354

Dear Sir:

In compliance with Section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating  ?

statistics for December are being forwarded to you with the summary l of changes, tests, and experiments that were implemented during-l l December 1993 pursuant to the requirements of-10CFR50.59(b). l r

Sincerely yours, 1

))\JIHoveh R.

l General Manager -

Hope Creek Operations DR:ws:JC r Attachments  ;

C Distribution 1

i i

21.00G8 - ,

4 The Energy People p 9401240140 931231 [ ""*'#***

l PDR ADDCK 05000354 4 R PDR a. , , . , , . -

i

INDEX l

3 NUMBER SECTION OF PAGES  ;

Average Daily Unit Power Level. . . . . . . . . . .- 1 Operating Data Report . . . . . . . . . . . . . . . 3 Refueling Information . . . . . . . . . . . . . . . 1

Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 4 l

l l

l l

l

~l i

1

. . _ . . ~. . - -- - - -. ..- - --

~i i

OPERATING DATA REPORT ,

DOCKET NO.- 50-354 .

UNIT Hooe Creek j DATE 1/10/94 l COMPLETED BY V. Zabielski i TELEPHONE -(609) 339-3506 OPERATING STATUS i

1. Reporting Period December 1993 Gross-Hours in~ Report Period 744

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 i' Month ~ Date' Cumulative

5. No. of hours reactor was critical 632.4 8567.4: 52823.0

6. Reactor reserve shutdown hours 10.0 0.0 0.0L

7. Hours generator on line 611.7 8527.6~ 52032.5 ,

8. Unit reserve shutdown hours 0.0 0. 0 - 0.0 .

9. Gross thermal energy generated 1975823 '27750151 165963370 (MWH) i

10. Gross electrical energy 665100 9215900 54963954 generated (MWH)  !

11. Net electrical energy generated 635975 8825300 52527684 '

(MWH) ,

12. Reactor service factor 85.0 97.8 85.7

13. Reactor availability factor 85.0 97.8 65,7

14. Unit service factor 82.2 97.3 H433.

15. Unit availability factor 82.2 97.3 84,,4

16. Unit capacity factor (using MDC) 82.9 97.7 82.6

17. Unit capacity factor 80.1 94.4 79.S (Using Design MWe)

18. Unit forced outage rate 17.8 2.7 41 2

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

RFOS scheduled for March 1994

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

N/A  ;

1 i

i i

_. - .J

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

UNIT HoDe Creek DATE 1/10/94 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506

-i MONTH December 1993 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR l F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (HOURS) (1)_ POWER (2) ACTION / COMMENTS 1 12/1 F 132.3 A 2 Unit shut down due-to excessive arcing of the main generatoriexciter ,

brushes.

• 2 12/17 F 0 A 5 Unitipower reduced to enable replace-ment of failed condenser tubesheet plug.

I s

i

. 1 1

AVERAGE DAILY UNIT POWER LEVEL i i

DOCKET NO. 50-354 i UNIT Hope Creek DATE 1/10/94 i COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 MONTH December 1993 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 15 17. 909

2. p 18. 1073

3. 2 19. lagg

4. 2 20. 1064

5. A 21. 1064

6. 146 22. 107_9_

7. 1021 23. 1068 i

8. 1060 24. 1064  !

9. 1060 25. 1065

10. 1073 26. 1065

11. 1056 27. 1072

12. 1053 28. 1071 f

13. 1068 29. 1072

14. 1064 30. 1068

15. 1056 31. 1068

16. 1043 1

i 1

l REFUELING INFORMATION DOCKET NO. 50-354 UNIT Hope Creek 1 DATE January 8. 1994 COMPLETED BY S. Hollinasworth TELEPHONE (609) 339-1051 ,

MONTH December 1993

1. Refueling information has changed from last month:

Yes No X

2. Scheduled dre3 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:

Not scheduled vet. ,

6. Important licensing considerations associated with refueling:

!!1h

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage (prior to refueling) 1008 C. In Spent Fuel Storage (after refueling) 1240

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 (EOC13) ,

licensed capacity: '

(Does allow for full-core offload)

(Assumes 244 bundle reloads every 18 months until then)

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

i I

HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

December 1993 Hope Creek entered the month of December at 100% power. The unit was manually shutdown at 0050 on December 1, 1993 due to excessive sparking from the main generator excitor brushes. The unit was restarted on December 6, 1993. It continued to operate through the end of the month. A brief power reduction was performed on December 17 to allow taking a circulator out of service to replace a leaking tubesheet plug. As of December 31, the plant has been.

on line for 25 consecutive days.

h

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION December 1993 The following items have been evaluated to determine:

1. If the probability of occurrence cnr~ 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 basisfforiany.

technical specification is reduced.

The 10CFR50.59 Safety Evaluations showedithat 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.cn: environmental questions are involved.

l i

.j

) -

i.

2 I

i QCE - Summary of Safety EvaluatiQD 4EC-3407 This Design Change Package completed the l installation of the reactor level '

Package 2 instrumentation required by NRC Bulletin 93- ,

03. Package 1 (reported at earlier date) installed all external non-intrusive

. equipment. This package provided the engineering analysis, design and instructions-  :

, to make the interface connections between the '

CRD and reactor. vessel level, systems._ The '

test instructions included functional and pressure testing of the system and demonstrated reliable operation of the connected transmitters when transients occur, j The loss of back-fill failure does not prevent e the mitigation of accidents by these '

t 7smitters. All automatic protective ac. ations utilizing the level. instruments on r the reference leg connected to the backfill system will initiate before the reactor has

'depressurized to a point where rapid expulsion ,

of noncondensables from~the reference leg could cause the level transmitters to sense a false high' level.

Therefore, this DCP does not increase'the~

probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed-Safety Question. '

Procedure Summary of Safety Evaluation NC.NA-AP.ZZ-0001 E These two Nuclear Department Administrative NC.NA-AP.ZZ-0032 Procedures govern the various aspects of the procedure control program at' Artificial Island. .They were revised to update and  ;

clarify the definition of 'shall' and 'should' for their use in Nuclear Department procedures.

Therefore, these Procedure revisions do not

' increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question. .

NC.NA-AP.ZZ-0007, These four Nuclear Department Administrative i NC.NA-AP.ZZ-0024, Procedures govern ALARA, Radiation Protection )

NC.NA-AP.ZZ-0029,& '

NC.NA-AP.ZZ-0045 Program,iratory and Resp Protection ProgramRadioactive Material Controlj respectively. -All of these procedures have i been updated to comply with the new 10 CFR 20 rules.

i

i Therefore, these Procedure revisions do not increase the probability or consequences of an ;

accident previously described in the SAR and i does not involve an Unreviewed Safety Questio.1.

NC.NA-AP.ZZ-0025 This Nuclear Department Administrative Procedure governs the Fire Protection Program.

This revision incorporates modifications to the Hot Work Permit for specific instructions for work on or over gratings and floor openings. This procedure change does not affect the evalaations on fire protection equipment or their performance, nor does it change criteria or assumptions used to develop the Fire 1:azarri Analysis.

Therefore, this Procedure revision does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.

Temocrar? Summary 91 Safety Evaluation '

Modifications93-026 This Modification temporarily installed jumpers for #2 Feedwater heater Hi/Hi level switches during low power operation which could cause false hi level signals to the Feedwater Heaters. This modification was removed on December 7, 1993 after the restart was completed. This Temporary Modification does not affect the normal and high level controls. Failure of the Feedwater system does not compromise any safety.related system or components or prevent the safe shutdown of the plant.

Therefore, this Temporary Modification-does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.93-028,029,030, These temporary modifications to the "A" ,

& 031 throuch "D" Traveling Screens allow continuous i operations of the screens and spray wash pumps  !

during low speed operations when the control ,

switch is in " AUTO". The modification is a '

jumper around the timer which prevents the ,

screen from starting until a preset differential pressure has been reached.

1

.syvy

-- T

l Failure of this T-Mod would cause the' circuit  ;

to shift back to original design. If.the screens should stop, there is a head shaft indicator feeding the CRIDS display which will alert operations.

Therefore, this Temporary Modification does not increase the probability or consequences '

of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.

QiLid'ency Report Summary of Safety Evaluation i

EMD v3-059 This Deficiency report dispositions the repair of the failure of the Silicon Controller Rectifiers (SCR) master and slave controllers for the Control Room Supply (CRS) unit heater circuitry. The disposition will allow the l repair of the Master SCR with a new spare SCR I power section due to limited parts inventory.

l It will remove two of the four SCR which will j reduce the design heating requirement to j approximately 50% capacity, from 90Kw to 45Kw.

, As discussed in the UFSAR Table 9.4-2, a postulated failure of the electrical heating coils from the CRS-unit has no affect on control room ventilation system during a DBA.

The electrical heater coils are not required to operate during emergency operation.

l Therefore, this Deficiency Report does not l l increase the probability or consequences of an accident previously described in the SAR and  ;

does not involve an Unreviewed Safety '

Question.

HMD 93-061 This Deficiency Report dispositions an elbow on the steam seal evaporator continuous blowdown line which was leaking. A temporary leak repair on the 1" 90* elbow was installed.

The proposed disposition will encapsulate the elbow with a clamp and inject leak sealant compound (if required) to minimize or stop steam leakage.

The steam seal system does not have any safety related function. Its failure does not affect l

l any safety-related system or component or prevent the safe shutdown of the plant.

l l

t i

l l

l

• • Therefore, this Deficiency Report'does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.

Other Summary gi Safety Evaluation [

UFSAR Sect 10.4.6 This Safety Evaluation covers revisions to I Change Section 10.4.6 (Condensate' Cleanup System) of H1-AK-NSE-0830 the HC UFSAR. Experience with the system has shown that various operating criteria 3 established at the time this section was

-written no longer reflect the optimum manner in which the system can-be operated. The changes include different cation to anion resin ratio, clarification of resin .

l regeneration / cleaning methodology, deletion ~of '!

several on-line effluent analyzers. j The Condensate Demineralizer System provides i no safety function._The changes to the UFSAR  :

do not affect the design: intent of the system .;

to provide high purity _feedwater to the l reactor. < ,

Therefore, this UFSAR_ change.does not increase' ,

the probability or_ consequences of an accident; '

previously described in the SAR and does not .

involve an Unreviewed Safety Question.  !

l l

l 1

l I

I l

.. . - .-. .