ML20082J652

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

ML20082J652
Person / Time
Site:	Hope Creek
Issue date:	03/31/1995
From:	Hovey R, Lyons D, Schmidt R Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9504180339
Download: ML20082J652 (13)

Text

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' O PSEG-Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station April 14, 1995 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 March are being forwarded to you with the summary of changes, tests, and experiments that were implemented during March 1995 pursuant to the requirements of 10CFR50.59 (b).

S cerely yours, fw&

J. Irovey General Manager -

Hope Creek Operations ik/l &

DR:WS:JC Attachments C

Distribution 1 CC10 *r i

The Energy People

/

9504180339 950331 S,,n g.,ygg.,

PDR ADOCK 05000354 R

PDR

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e INDEX i

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NUMBER i

SECTION OF PAGES Average Daily Unit Power Level.

1 Operating Data Report.

3 l

Refueling Information.

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l l

Monthly operating Summary.

1 j

l Summary of Changes, Tests, and Experiments.

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4 9

OPERATING DATA REPORT DOCKET NO.

50-354 UNIT Hope Creek DATE 04/05/95 COMPLETED BY D.

W.

Lvons

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TELEPHONE (609) 339-3517 OPERATING STATUS 1.

Reporting Period March 1995 Gross Hours in Report Period 111 l

2.

Currently Authorized Power Level-(MWt) 3293 Max. Depend. Capacity (MWe-Net) 1031 Design Electrical Rating (MWe-Net) 1067 3.

Power Levci 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 584.6 2000.6 61936.5

)

i 6.

Reactor reserve shutdown hours 222 922 22Q 7.

Hours generator on line 563.0 1979.0 60982.4 i

8.

Unit reserve shutdown hours 0.0 222 922 1

9.

Gross thermal energy generated 1781131 6398945 194813291

)

(MWH) j

10. Gross electrical energy 593774 2147162 64574829 generated (MWH)

11.. Net electrical energy generated 566874 2057037 61710353 (MWH)

12. Reactor service factor 78.6 92.6 85.3 3

1

13. Reactor availability factor 78.6 92.6 85.3

14. Unit service factor 75.7 91,6 84.0

15. Unit availability factor 75.7 91.6 84.0 l

16. Unit capacity factor (using MDC) 2222 92.4 82.5

17. Unit capacity factor 71.4 89.3 79.7 l

(Using Design MWe)

18. Unit forced outage rate 24.3 121

_122

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

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OPERATING ~ DATA REPORT' UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.

50-354 UNIT Hone' Creek

'DATE 04/05/95 COMPLETED BY D.

W.

Lyons TELEPHONE. (609) 339-35 MONTH ' March 1995 METHOD OF i

SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.

DATE S= SCHEDULED (HOURS)

(1)'

POWER (2)

ACTION / COMMENTS 1.

3/18-OTHER 0

H 5

A FIRE UNDER THE 3/19 KEENEY ELECTRIC TRANSMISSION LINE NECESSITATED TAKING THAT LINE OUT OF SERVICE.

WITH THE KEENEY LINE-OUT OF

-i SERVICE, IT WAS NECESSARY TO REDUCE j

POWER AT HOPE CREEK DUE TO GRID STABILITY CONCERNS.

INDUSTRY GUIDELINES FOR OUTAGES USUALLY DO NOT CLASSIFY THIS TYPE OF EVENT AS EITHER A FORCED OR SCHEDULED OUTAGE.

l 2.

3/19-FORCED 0

A 5

FAILURE OF A POTHEAD 3/20 ON THE 1AX501 XFRMR l

REQUIRED THE I

OPERATORS TO REDUCE POWER.

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4 OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS i

DOCKET NO.

50-354 i

UNIT HoDe Creek DATE 04/05/95 COMPLETED BY D.

W.

Lyons TELEPHONE (609) 339-351 i

MONTH March 1995 1

l METHOD OF l

SHUTTING DOWN THE 4

TYPE REACTOR OR i

F= FORCED DURATION REASON REDUCING CORRECT.IVE NO.

DATE S= SCHEDULED (HOURS)

(1)

POWER (2)

ACTION /COMr4ENTS 3.

3/20-FORCED 181 G

2 WHILE I&C TECHS WERE 3/27 PERFORMING A PM PROC ON THE OPTICAL l

ISOLATOR FOR THE REACTOR RECIRCULATION PUMP i

MG SETS THE OPERATOR EXPERIENCED A LOSS OF BOTH MG SETS. HE THEN INITIATED A MANUAL SCRAM IN l

ACCORDANCE WITH THE PROCEDURE.

THE LOSS OF THE MG SETS WAS CAUSED BY THE TECH'S OPERATION OF THE DVM i

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AVERAGE DAILY UNIT POWER LEVEL I

DOCKET NO.

50-354 UNIT Hone Creek DATE- 04/05/95 COMPLETED BY D.

W.

Lyons

{

TELEPHONE (609) 339-35' l

MONTH March 1995 i

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

(MWe-Net) i 1.

1059 17.

1047 2.

1063 18.

121 3.

1064 19.

15Q 4.

1063 20.

lag j

5.

1050 21.

A i

6.

1058 22.

A 7.

1053 23.

A i

8.

1050 24.

2 9.

1059 25.

R l

10.

1066 26.

2 I

11.

1055 27.

2 12.

1055 28.

157 j

13.

1050 29.

1054 14.

1051 30.

1055 15.

1050 31.

1061 16.

1048 i

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REFUELING INFORMATION DOCKET NO.

50-354 UNIT Hooe Creek 1 DATE 94/05/95 COMPLETED BY R.

Schmidt TELEPHONE (609) 339-374-

' MONTH March 1995

-l 1.

Refueling information has changed from last month:

1 Yes X

No j

2.

-Scheduled date for next refueling:

11/11/95 3.

Scheduled date for restart following refueling:

12/10/95 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?

Auaust 28. 1995 5.

Scheduled date(s) for submitting proposed licensing action:

HQt reauired.

6.

Important licensing considerations associated with' refueling:

)

HIA i

7.

Number of Fuel Assemblies:

i A.

Incore 211 B.

In Spent Fuel Storage (prior to refueling) 1240 C.

In Spent Fuel Storage (after refueling) 1472 l

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Present licensed spent fuel storage capacity:

4006 I

Future spent fuel storage capacity:

4006 j

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 n21 allow for smaller reloads due to improved fuel) t t

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

March 1995 The Hope Creek Generating Station began the month'on-line at 100%

power and remained there until a marsh fire under the Keeney (5015) Electric-Transmission line required that'line to be taken out of service.

Because of the line outage, unit power was-reduced in accordance with grid stability curves from 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on March 18, 1995 until 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on March 19, 1995.

' Unit power was reduced at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on March 19, 1995 when a pothead insulator on the 1AX501 Station Power Transformer failed.-

This power reduction was ended at 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br /> on March 20, 1995 when the unit was manually scrammed in accordance with procedures upon the loss of both' reactor recirculation pumps.- I&C Technicians were performing a PM Check an the optical isolator for.

the Reactor Recirculation Pump Motor Generators.

The Technician's handling of the digital multi-meter (DMM) caused the. loss ~of_the MG sets.

Power operation resumed again when the reactor-was taken critical at 0019 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> and returned to service when the generator was synchronized to the grid at 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br /> on March 27, 1995.

Full power was achieved at 1956 hours0.0226 days <br />0.543 hours <br />0.00323 weeks <br />7.44258e-4 months <br /> on March 28, 1995.

Hope Creek ended the month at 100% power and has been on-line for four days.

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION March 1995 The following items have been evaluated to determine:

1.

If the probability of occurrence or the consequences of an accident cnr malfunction of equipment important to safety i

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 i

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 i

environmental impact.

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

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j Desian Chances Summary 21 Safety Evaluation l

4EC-3394 Egg 11 This design Change Package installed a 4 inch-return line manual gate valve on the Reactor Water Clean Up-(RWCU)in the to the Feedwater "B" Header.

The Valve was installed locked open position upstream of the existing check valve in l

Reactor Building Room 4316 at Elv. 102'.

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i This addition will allow the continuous use of RWCU while each Feedwater line can be individually tested (LLRT etc.).

Currently i

l the RWCU System must be removed from service during Feedwater Header LLRT resulting in decreased water clarity and higher refueling floor dose rates during refueling outages..This modification does not adversely compromise any safety related system or component or prevent the safe shutdown of the plant.

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 Summarv af Safety Evaluation Administrative Procedure NC.NA-AP.72-0012 Eay 11 This' Nuclear Department Administrative Procedure is titled Technical i

Specifications' Surveillance. Program.

This change' modifies a commitment made to the NRC by Sa?.em Station in response to a l

Notice of Violation (NLR-N89176).

It required SORC review of station readiness to implement Tech Spec Amendments prior to implementation.

This SORC review was established in 1989 as the last stage of a comprehensive formal review process for Tech Spec Amendment implementation.

This systematic process involves a review by each station department to identify all documents that require revision in accordance with the. Tech spec Amendment.

As the departmental review process at Salem has evolved it has since gained more thoroughness and efficiency.

The last stage of the review process is the SORC review.

It is the consensus of the

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SORC that their review adds little to the amendment preparation process.

It will therefore be eliminated at Salem Station in favor of a simple notification to station management that the primary review process has been successfully completed prior to j

the Tech Spec Amendment implementation.

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.

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1 Temocrary Modification Summary,pf EAfety Evaluation T-Mod 95-008: This Temporary Modification installed a blank in l

place of flow orifice OHBFO-5313.

This is needed due to seat I

l leakage across valve OHBHV-F049.

This valve would normally l

isolate the waste-sludge phase separator (WSPS) from the equipment drain filter processing flow path.

The degraded valve caused the WSPS to become full forcing removal of the waste filter from service.

The Purpose of the orifice along with the F049 valve was l

l to limit the flow from the waste collector and waste surge tanks l

.ta) the WSPS during tank blowdowns.

There are no credible failure modes associated with this change.

The blank is sized IAW Power Piping Code B31.1 criteria and requirements for blanks.

The proposal does not affect any l

pressure retaining characteristics of the piping and has no effect on the ability of the system to contain radioactive material.

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.

T-Mod 95-010: The Torus Level transmitter has lost the use of it's reference leg which senses suppression pool atmosphere due to isolation of valve 1BJHV-4865.

In order ta) correctly _ indicate torus water level at the Remote Shutdown Panel alternate indication for this parameter, the valve lineup for the indicator must be changed.

The resulting change vents the low pressure reference tap to the surrounding Core Spray Room.

An existing Q transmitter will be used to provide suppression pool atmosphere

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pressure readings that will serve as a correction factor which when subtracted from reading obtained, will accurately represent torus water level.

l Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in j

the SAR and does not involve an Unreviewed Safety Question.

l T-Mod 95-012:.This Temporary Modification will perform a temporary repair on Regen Waste transfer line.

This below grade pipe was damaged during excavation near the new Services Building.

This pipe is used to transfer waste water from the Regen Waste tank to the non-radwaste treatment basin at Salem Station.

The repair will be accomplished using 6" PVC pipe and couplings until a

. permanent repair can be made using ABS as specified on note'10 on P&ID M-18-0 sht 1 (UFSAR Fig 9.2-7).

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.

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T-Mod 95-014: This Temporary Modification installed an electrical jumper across the #2 Feedwater Heater Hi-Hi Level trip switches and installed a temporary keep fill line to the low side of the level transmitters.

This modification is performed due to spurious indications during power ascension and is removed at approximately 40 % Reactor Power.

This T-Mod does not increase the probability or the consequences of an accident listed in Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resulting in a turbine trip.

i 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.

Deficiency Report Summary of Safety Evaluation DEL 950223135: This Deficiency Report documents the plugging of tubes in the RACS Heat Exchanger up to a total of 50.

This resulted in a change to the heat removal capacity of that Heat Exchanger as described in the UFSAR Table 9.2-9.

The plugging of up to 50 tubes is within the analyzed change allowed per Nuclear Engineering Assessment NE-95-0156.

This analysis reviews the heat exchangers for a worst case of 50 tubes plugged per heat exchanger.

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 Chance Notice 95-01:

Hope Creek Chemistry requested that passageway 3439 located in elv 124' in the Services and Radwaste Building be reviewed for storage of files and other office support equipment.

The Calculation determined that the fire load is increased by 1.29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 17 min).

The original load was 10 min. (from table 9a-1).

This increase provides no distinguishable increase to the total fire severity of this fire area (AB3).

This safety evaluation supports both the additional of this combustible material to the passageway and to update the combustible lad data i

on Table 9A-1 and 9A-6 of Appendix A of the UFSAR.

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Based on General Criteria 9A.1.2 the spread of fire is assumed to be limited to the fire area in which the fire originates due to l

separation from other fire areas by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barriers.

j Therefore, this UFSAR Change notice does not increase the i

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

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a UFSAR Chance Notice "Section 13; Conduct of Operations, Section 17.2; Quality Assurance During the Operation Phase, and QA/NSR Procedures and functions described in the HC UFSAR":

A reorganization of the Quality Assurance (QA) and Nuclear Safety Review (NSR) Departments have occurred.

The reoganization of QA affects only personnel titles with respect-to applicable Technical Specification sections 6.5.1.8, 6.5.1.9, 6.5.2.7 and section 17.2 of the HC UFSAR.

There is.no reduction of QA functions described within the references.

The same Tech Spec title changes described above apply to the NSR Reorganization.

Similar title changes apply to UFSAR 13.1.1.2.1.4 and 13.4.4.

Section 13.1.1.2.1.4 contains the following commitment to the Human Performance Enhancement System (HPES): "The HPES Group administers the Human Performance Enhancement System, an INPO sponsored program which focuses on evaluating and improving human performance in Nuclear Power Plant Operations."

sponsored by INPO, is not a safety related program.

These HPES,iques have been incorporated in the NBU procedure-NC.NA-techn BP.ZZ-0002(Z) Root Cause Analysis Guidelines.

This procedure.

serves as a guide to all investigators of events using root cause analysis.

These techniques will have a wide spread use in the Nuclear Business Unit rather than an exclusive role of a centralized group.

Thus, deletion of the above commitment from the UFSAR will have no detrimental effect on nuclear safety.

Therefore the deletion of the HPES Group from the Hope Creek UFSAR will have no impact to safety.

Therefore, this UFSAR Change notice 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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