ML20081D686

Monthly Operating Rept for Feb 1995 for Hope Creek Generating Station Unit 1

ML20081D686
Person / Time
Site:	Hope Creek
Issue date:	02/28/1995
From:	Hovey R, Lyons D, Schmidt R Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9503200363
Download: ML20081D686 (10)

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O PSEG .

Pu,blic Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station i

! March 15, 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 l

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Dear Sir:

MONTHLY OPERATING REPORT '*

HOPE CREEK GENERATION STATION UNIT 1 DOCKET NO. 50-354 l

In compliance with section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating statistics for l February are being forwarded to you with the summary of changes,.

l j tests, and experiments that'were implemented during February 1995 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, L

i R. J. Hovey l General Manager -

Hope Creek Operations. I 9"W$b DR:WS:JC l

Attachments C Distribution 210000 J.

The Enerav Peoole /y '

9503200363 950228 L PDR ADOCK 05000354 es 2ir3 <23u> i2 89 R PDR

.' INDEX ,

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1 NUMBER I SECTION 'OF PAGES-1 Average Daily Unit' Power Level. . . . ....-...;.= 1- l l

Operating Data Report . . ... . . . ._. . .. . . . . . -

3 l Refueling Information ... ... ' . . . . . . . . . . . - .. . . 1:

Monthly operat.ing Summary:. . . . .. .... . . . .:.. . -1 Summary of Changes, Tests,.'and Experiments...

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• OPERATING DATA REPORT DOCKET NO. 50-354 UNIT. Hope' Creek DATE 03/02/95 jd i COMPLETED BY D. W. Lvons I"d '

TELEPHONE (609) 339-3517 i

OPERATING STATUS i

1. Reporting Period February 1995 Gross Hours in Report Period 111 a 2. Currently Authorized Power LevelE(MWt) 3293 i Max. Depend. Capacity (MWe-Net) 1031

Design ~ Electrical Rating (MWe-Net) 1067

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

4. ReasonsLfor restriction;(if any)-

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'This Yr To Month Date Cumulative-5 .~ No. of hours reactor was critical. 672.0 1416.0 61351.9

6. Reactor reserve shutdown hours. 222 222' 222

7. Hours generator on line 672.0 1416.0 60419.4 '

8. Unit reserve shutdown hours 229 Q2Q 222

9. Gross thermal energy generated 2207584 4617814 193032160 (MWH)

• - 10. Gross electrical energy 742042 1553388 63981055 generated (MWH)

. 11. Net electrical energy. generated .711976 1490163 61143479 j (MWH) i 1 12. Reactor service factor 100.0- 100.0 85.4

• 13. Reactor availability factor 100.0~ ^100.0 85.4

14. Unit service factor 100.0 100.0 84.1 'i

15. Unit availability factor 100.0 .100.0 84.1

16. Unit capacity factor (using MDC) 102.8 102.1 82.6

17. Unit capacity factor 99.3 98.6 79.8 (Using Design MWe).

18. Unit forced outage rate 222 222 4.6 I

19. Shutdowns, scheduled over next 6 months.(type, date, T u tion) :

None ~

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

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

DATE 03/02/95. ~ 4gb COMPLETED BY- D. W. Lvons TELEPHONE: (609) 339-3517 MONTH- Februarv 1995 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. NONE

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.' . AVERAGE DAILY' UNIT POWER. LEVEL DOCKET NO. 50-354 UNIT Hooe Creek DATE 03/02/95 COMPLETED BY~ D. W.-Lyons TELEPHONE (609) 339-3517-MONTH February 1995 DAY AVERAGE DAILY POWER-LEVEL DAY AVERAGE-DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 1061 17. 1057'

2. 1062 18. 19.11

3. 1065 19. 1053

4. 1062 20. 1055

5. 1059 21. 1069

6. 1065, 22. 1063

7. 1048 23. 1053

8. 1061 24. 1Q11

9. 1066 25. 1061

10. 1057 26. 1055 i

11. 1060 27. 1064 J

12. 1053 28. 1058

13. 1059 29. n/a 4

14. 19.11 30. n/a

15. 1060 31. n/a..

16. 1062 I

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.' REFUELING INFORMATION 1 l

DOCKET NO. 50-354 UNIT. Hooe Creek 1 '

i DATE 03/02/95 COMPLETED BY Schmidt R.

! TELEPHONE (609) 339-3740

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l MONTH Februarv'1995

1. Refueling information has changed from last month:

l Yes X No

2. Scheduled date for next refueling: :11/09/95

3. Scheduled date'for restart following' refueling: 12/09/95

4. LA. Will Technical Specification changes or other license l amendments be required?

Yes No X B. Has the Safety Evaluation covering the COLR been reviewed by the *

[ Station Operating. Review Committee?

l Yes No X l If no, when is it scheduled? Auaust 28, 1995 5.. Scheduled date(s) for. submitting proposed licensing action:

Upt reauired.

6. Important licensing considerations associated-with refueling:

Hus

7. Number of Fuel Assemblies:

A. Incore . 25.i B. In Spent Fuel Storage (prior to refueling) 1240 C. .In Spent Fuel Storage (after refueling) 1472

8. Present licensed spent fuel storage capacity: 40Q5 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 Dgt allow for smaller reloads due to improved fuel) l l

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

February 1995 8

Hope Creek entered the month of February operating at 100% power.

It continued to operate at 100% power throughout the month and as of February 28 1995, the ur.it has been on line for 140 consecutive days.

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SUMMARY

OF CHANGES, TESTS,.AND EXPERIMENTS, i i . .

i FOR THE HOPE-CREEK GENERATING! STATION

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JFebruary 1995.

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-1 lThe_followingfitems have been; evaluated.to' determine: ,

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1. If'the probability of occurrence or the consequences of an-

accident or malfunction.of equipment 11mportant:to; safety f l l . previously evaluated in_the safety analysis: report may M '

increased; or

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2. If ~a possibility for an ' accident or malfun'ction of a' different ,

type than.any evaluated previously'in thetsafety analysis-  ;

report.may be' created; or.  ;

3. . If - the margin of safety :as defined f in the basis f for' any I technical specification.is reduced.

The 10CFR50.59 3afety Evaluations showed that these items did'not-create a new.ief-Ly hazard to the' plant nor did they affect the j

safe shutdown 4 :he reactor. These items did notechan plant effluent reiaases and did not' alter the. existing.ge the '

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environmental. impact. The 10CFR50.59 Safety Evaluations . .

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, determined that no unreviewed safety or environmental questions

! are involved.

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Summary 21 Safety Evaluation  !

f ]pesidnchanaes-

] 453-0144: This Design Change will add a sample point for'the RCIC -

Turbine Lubrication Oil-System. A valve will be installed in a

, spare port in the piping cross fitting on-the. oil drain line from

the coupling and bearing. The addition of a sample valve will j facilitate more consistent and. representative

oil samples.and-j_ decrease the time. required.to obtain them (ALARA). 'Also, an:

) inaccuracy that was found in the'RCIC system ~P&ID during.the J review of the subject change was also' corrected (the-information i pertaining to the RCIC Bearing oil thermometers and.high a

temperature switch was incorrect). The correct information was j obtained from the Vendor Drawing and incorporated.

h Since the new valve will.be connected to what'is-currently'a-spara 4 port on a pipe cross fitting, it.will have no~effect on flow in j .the drain pipe or operation of the thermometer which is connected.

to the opposite port. .The valve will~be Q-rated' It j

will be installed and locked.in the closed positi_ on, Seismic opened 1.onlyz j to. draw a sample.

It will have:no impact'on the' operation of the

RCIC Turbine or associated instrumentation. 'Therefore, there is

no credible failure modes associated with this. change. ]

'Therefore,'this DCP does not increase the probability or i consequences of an accident previously described in the SAR and i . does notl involve any Unreviewed Safety Question.

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[ Temoorary Modificqhi2D Summarv'gf Safety Evaluation i 95-006: This temporary modification installed five (5) mechanical i

plugs deviations in five (5) identified. Reactor The where Building drain lines'when deviations were documented design on i Incident Report 950131205 following an Operating; Experience review 1 by Nuclear Engineering of-a Nine Mile Point LER (94-006) which

identified an issue with the secondary containment.

! The Decontamination Radwaste and Dirty Radwaste header

{ configurations were reviewed for potential containment integrity 1 a - concerns due to a leakage path through the-floor drains. The i investigation showed no evidence of: water-loop seal-in the various i i floor drains that ties both the Reactor Building and the Turbine l Building floor drains tWether. These drain plugs will secure 2 - potential leakage ~ paths in lieu of water seals as is currently j stated in the UFSAR. j 4: .

v There is no credible failures associated with the installation of -

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, plugs in lieu of the water loop seal as far as secondary s containment is concerned. The capability of the FRVS and RBVS

systems in maintenance of the required secondary containment i negative pressure is not impacted by this modification. 4 Radiological consequences are not altered due to

the fact _that' both RBVS'and FRVS systems have been~able-tb-maintain the required p pressure in the secondary containment.

Therefo?e, this Temporary Modification-does not increase the l probability or consequences of an accident previously-describedfin j the SAR and does not involve an Unreviewed. Safety Question.

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.Other Summary 21 Safety Evaluation LCO 95-150t'This safety evaluation addresses the continued .

i functionality of the "A" Emergency Diesel Generator while one'(1) of the Fuel Oil Transfer. Pumps is inoperable. Hope. Creek Tech .

Specs (3.8.1.1.b.3) requires each of the four. diesel generators to-

be supported by a separate fuel oil transfer pump for each.of-its

! two fuel oil storage tanks. Each' fuel oil transfer pump is a 100% ,

! capacity pump with its own piping and controls capable of- ,

supplying the' day tank to support the fuel oil supply for its assoclated diesel.

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i In order to justify the continued-functionality of the EDG-operation UFSAR 9.5.4.2.2 was referred to. This states.that the-purpose of providing two (2) pumps.is to " equalize" the wear.

Each pump is capabla of delivering 43 gpm to-the day tank. The diesel consumption at 100%. load is 5.8.gpm. 'Because the capacity of the fuel oil transfer is greater than the fuel consumption of a the-diesel engine.it can supply oilito the engine and

,' simultaneously increase the inventory of the day tank.

Therefore, this Safety Evaluation does.not-increase the . .

probability or consequences of an accident previously described in

_ the SAR and does not involve an Unreviewed Safety-Question.-

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i H-1-RC-MSE-0843:. The' purpose of this Safety. Evaluation is to identify systems structures _and components-(SSC).and justify <their position as being outside the scope of Nuclear Jurisdiction. The ,

Design Change Process (DCP) would not apply to modifications to

! these SSC's and create a DCP Exclusion Zone. This also removes i P&ID M-23-1 Sheet 1 From UFSAR (Fig. 9.3-4) per SAR Change 94-43.

, This Exclusion Zone will encompass the Turbine Building Sample 4 Station.

k Any tubing failures inside the panel will be. bounded by'the l Feedwater pipe break outside of containment UFSAR Section 15.6.6

} and Instrument Line Break UFSAR Section 15.6.2. These have 1

marginal applicability only because the panel draws a sample from

the main steam line drains. There~1s no equipment important to

! safety within the DCP Exclusion Zone boundary. Any equipment'that communicates-with the exclusion zone either electrically or i mechanically is either not important to safety or the failure has i been previously evaluated.

i' Therefore, this Safety Evaluation does not increase the probability or consequences of an accident previously. described in '

the SAR and does not involve an Unreviewed Safety Question.

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