ML20078A947

Monthly Operating Rept for Dec 1994 for Hope Creek Generation Station Unit 1

ML20078A947
Person / Time
Site:	Hope Creek
Issue date:	12/31/1994
From:	Hovey R, Lyons D Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9501250199
Download: ML20078A947 (14)

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T "o psgc Public Service Electric and Gas Company P.O. Box 236 Hancocks Bndge, New Jersey 08038 Hope Creek Generating Station January 13, 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 December are being forwarded to you with the summary of changes, tests, and experiments that were implemented during December 1994 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, R. J. Hovh General Manager -

Hcpe Creek Operations DR:WS:JC Attachments C Distribution 2400Cli The Eneray People '

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9501250199 941231 '

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PDR ADOCK 05000354 R PDR

INDEX NUMBER SECTION OF PAGES Average Daily Unit Power Level. . . . . . . . . . . 1 Operating Data Report . . . . . . . . . . . . . . . 3 Refueling Information . . . . . . . . . . . . . . .. 1 Monthly Operating Summary . . . . . . . . . . . . . 1 "

t Summary of Changes, Tests, and Experiments. . . . . 6 1

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OPERATING DATA REPORT DOCKET NO. 50-354 UNIT HoDe Creek DATE 01/05/95 COMPLETED BY D. W. Lyons TELEPHONE (609) 339-3517 OPERATING STATUS

1. Reporting Period DECEMBER 1994 Gross Hours in Report Period 211

2. Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity 1031 Design Electrical Ratin(MWe-Net) g (MWe-Net) 1067 l l

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

4. Reasons for restriction (if any)

This Yr To Month Date Cumulative

5. No. of hours reactor was critical 744.0 7112.9 59935.9

6. Reactor reserve shutdown hours 212 0.0 Hzg

7. Hours generator on line 744.0 6970_2 59003.4 t

8. Unit reserve shutdown hours 0.0 0.0 0.0

9. Gross thermal energy generated 2430315 22450976 188414346 i (MWH)

10. Gross electrical energy 819216 7463712 62427666 generated (MWH)

11. Net electrical energy generated 786182 7125632 19653316 (MWH)

12. Reactor service factor 100.0 81.2 85.1 l

l 13. Reactor availability factor 100.0 81.2 85.1

14. Unit service factor 100.0 79.6 83,8

15. Unit availability factor 100.0 1215 83.8

16. Unit capacity factor (using MDC) 102.5 78,9 82.2

17. Unit capacity factor 99.0 76.2 79.4 (Using Design MWe)

18. Unit forced outage rate 212 122 Axl

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 l UNIT SHUTDOWNS AND POWER REDUCTIONS l

DOCKET NO. 50-354 UNIT HoDe Creek DATE 01/05/95 COMPLETED BY D. W. Lyons h '

TELEPHONE (609) 339-3517 l

l l MONTH DECEMBER 1994 METHOD OF SHUTTING  :

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

1. NONE '

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AVERAGE DAILY UNIT POWER LEVEL DOCKET No. 50-354 UNIT HoDe Creek DATE 01/05/95 COMPLETED BY D. W. LvonsAN' TELEPHONE (609) 339-3517 MONTH DECEMBER 1994 i

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

1. 1067 17. 1062

2. Aqii 18. 1054

3. 1063 19. 1063

4. 1058 20. 1063

5. 1052 21. 1060

6. 1056 22. 1061

7. 1053 23. 1057

8. 1064 24. 1050 l

9. 1069 25. 1057

10. 1045 26. 3059.

11. Rig 27. 1060

12. 1071 28. 1053 l

13. 1065 29. 1069

14. 1062 30. 1066

15. 1063 31. 1063

16. 1059 i

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l REFUELING INFORMATION i

l DOCKET NO.19-354 UNIT HoDe Creek 1 DATE Jan 05, 1995 jgf COMPLETED BY R. Schmidt r*

TELEPHONE (609) 339-3740 l

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! MONTH DECEMBER 1994 i

l 1. Refueling information has changed from last month:

l l Yes X No

2. Scheduled date for next refueling: 10/14/95 ,

3. Scheduled date for restart following refueling: 11/13/95

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

• ies No X l B. Has the Safety Evaluation covering the COLR been reviewed by the l Station Operating Review Committee?

Yes No X l

If no, when is it scheduled? Auaust 78, 1995

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

Egt recuired, i

6. Important licensing considerations associated with refueling:

HLA

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage (prior to refueling) 124Q C. In Spent Fuel Storage (after refueling) 1472

8. Present licensed spent fuel storage capacity: 4006 Futu , spent fuel storage capacity: 4006

9. Date of last refueling that can be discharged 5/3/2006 l to spent fuel pool assuming the present (EOC13)

! licensed capacity:

l (Rgen allow for full-core offload)

! (Assumes 244 bundle reloads every 18 months until then)

(Does Dpt allow for smaller reloads due to improved fuel) l l

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

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December 1994 l

Hope Creek entered the month of December operating at 100% power. I The unit operated at full power without any major power reductions l or scrams. A minor power reduction to 70% for Main Steam drain valve repairs was accomplished on December 10 and 11, 1994. As of December 31 1994, the unit has been on line for 81 consecutive days. l t

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS

, FOR THE HOPE CREEK GENERA'fING STATION t

i DECEMBER 1994 l

l The following items have been evaluated to determine:

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i 1. If the probability of occurrence or 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 fer 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.

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The 10CFR50.59 Safety Evaluations showed that these items did not 4 create a new safety hazard to the plant nor did they affect the l 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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Desian Chanaes Summary 91 Safety Evaluation-4EA-0041 Pkg. 1: This Design Change Package adds a new manually operated globe valve and funnel to the Cooling Tower Blowdown (CTBD suction of RMS Sample Pump. It is added to the. existing suction line-to facilitate the priming of the pump.. The design specifications used to establish the specified service conditions of the CTB RMS will not be degraded because the addition will enhance performance of the blowdown-sample cump.and the system preventing pump or system failure.

There are no changes to the Technical Specifications (TS) as a

-result of this DCP because there are no Hope Creek TS associated with CTB RMS. No design specifications will be degraded. NRC's prescribed operating limits provide sufficient operating range i such.that the acceptance limits are not exceeded during plant i operations and analyzed transients will not be affected. Since .)

the acco tance limits will not be exceeded, there is no impact on. i the marg n of safety.

l Therefore, this DCP does not' increase the probability or ,

l consequences of an accident previously described in the SAR and  ;

does not involve any Unreviewed Safety Question, j i

4EE-0044 Pkg. is This Design Change Package documents the addition of break flanges downstream of a manually operated gate i valve. This is to facilitate the draining of the condensate .

system durin refueling outages. The facility as described in the  !

SAR is chang ng because FSAR Fig. 10.4-5 Sht.1 will be updated l incorporating the flanges.  !

This DCP documents the addition of flanges to an existing drain  ;

line and therefore does not change the original design parameters 1 of the Condensate System. The specifict* ions used to establish service conditions of the Condensate S)1cem will not be degraded i because all the design specifications are adhered to as documented ,

in the DCP. I

! 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.

4BC-5343 Pkg is (Combined discussion with Incident Report 354/94-233) This Design Change Package was initiated based on a condition identified in Hope Creek Incident Report 354/94-233. It includes two separate changes.

In the current system, when a SSW spray wash booster pump is running and a high differential water level is sensed across the screens, the screens will start. The logic will be changed to start the screens whenever the spray wash pump is pumping.

The second enange of the DCP will electrically remove the flow ,

switch in the discharge line of the spray booster pump. The l l interlock associated will the switch will be changed. The current  !'

l configuration of the interlock requires flow in the lines for the j screens to rotate. The new configuration of the interlock is that 1

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auxiliary contacts from the MCC of the booster pump will be used as'a permissive to run the screens. If the breaker to the booster 1 l

pump is closed, the screens will have a permissive to run.

l This change affects the Station Service Water Syster which

supports the RACS and SACS as well as Circ Water make-up. However j the change does not affect the way or the ability of the SSWS to l meet its support function. Therefore, the change does not affect l other systems.

l There are two credible failure modes for this change. The first i is for the screen to fail to start when the Service Water Pump is I

in service. The second Failure mode is for the screens to fail to stop when the booster pump is taken out of service. In either case, a zero speed switch input to CRIDS will alert the operator i that something is wrong. Applicable action statements can be i entered.

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.

4KA-0053 Pkg. 1: This Design Change Package documents the installation of spray covers to the Service Water intake structure fish and debris troughs. This change will modify Figures 1.2-40 and 1.2-41 of Hope Creek UFSAR Section 1.2. The new spray covers l were installed on the Service Water intake fish and debris trough l to prevent the spray of water on to the floor and surrounding j equipment. The higher humidity caused by the spray also helped to accelerate the corrosion processes of various components in the Traveling Screen Room. These modifications do not create any new ,

credible failure modes. l 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.

i i 5EC-3030 Pkg 1: This design Change Package documents the addition

! of a Low Level Radwaste Facility (LLRSF) to provide a safe, secure space for temporary storage of low level radwaste generated by both Salem and Hope Creek Generating Stations during a period

commencing when access to operating low level radwaste disposal facilities becomes unavailable and ending when an approved disposal facility becomes available to Public Service Electric and Gas. For storage volume design purposes, this period is 5 years in accordance with current Nuclear Regulatory Commission regulatory guidelines and recommendations.

The new facility is functionally and physically isolated from all systems, structures, components and cables which could affect reactor safety or be used to mitigate the consequences of an accident. Therefore, the LLRSF is non-safety related and not important to safety.

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.

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• i Deficiency Report Summary of Safety Evaluation l l l i

DR HMD 93-060 This Deficiency Report documents a " Repair" l

! disposition which disconnected one of thirty six failed heater l elements in the Control Room Supply (CRS) unit heater panel. This reduced the design heating requirement from 90 KW to approximately l 87.5 KW. UFSAR Table 9.4-2 discusses a postulated failure of the 1 l electric heating coils from the CRS unit has no affect on control l l room ventilation system during a design basis accident. The l l electric heating coils are not required to operate during emergency operations.

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.

DR HTE-94-157 This Deficiency Report documents the electrically I backseating of a startup drain valve. Also discussed here is an

! identical condition on another startup drain valve. This change proposes to reduce the valves' packing leakage during the operating cycle by changing the position of the valve from normally closed to normally open allowing the valves to be backseated. The leakage is causing elevated temperatures within the steam tunnel and poses a threat to surrounding equipment. The valves in question are startup drain valves and serve no function at power operations >50%. The lines will be isolated by other i normally closed valves. The line feeds the main condenser as a l drain path. Any leakage would be drawn to the condenser due to l the vacuum.

These valves do not perform an isolation function and no credit is taken for their closure or leak tightness. They are not part of the IST program and are not LLRT valves. Failure of these valves in the open position would not create any challenges to the  :

important to safety systems due to the fact that the bypass line  !

around the valves is normally open.

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

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1 Other Summary of Safety Evaluation I Safety Evaluation H-1-EA-KSE-0842 Rev Os This Safety Evaluation discusses Station Service Water System (SSWS) Pump net positive I I

suction head (absolute) [NPSHa] which is being changed from 27

, feet to 34 feet. HC UFSAR section 9.2.1.5 currently stated 1

available NPSHa for the station SSWS Pumps is 27 feet at the design low-low water level. The original plant design basis for these pumps stated that the available NPSHa was 35 feet. During a design change for increasing the SSWS design temperature from 85'f to 95*f, calculation of the NPSHa was made to be 27 feet. The l error occurred as the result from the mis-application of an ]

! unnecessarily conservative atmospheric pressure (considering the 2

effects of a local tornado) in the calculation. The increases in SSWS pump stated NPSHa (34 feet) thus increases the margin of q safety related to pump cavitation /operablity.

Therefore, this Safety Evaluation does not increase the a

probability or consequences of an accident previously described in 1 the SAR and does not involve an Unreviewed Safety Question.

SAR Change Notice 90-13: This SAR Change Notice documents a change l 4

to the licensing commitment statement in the UFSAR Section 9.3.3.2 l l Item 2, which states "there are no areas in Hope Creek Plant which i

contains both radioactive and non-radioactive drains." As a i

result of some interdepartmental reviews of a previous SAR Change ,

Notice, which was associated with DCP's to re-route condensate I drains from air handling unit cooling coils from DRW to the normally nonradioactive Turbine Building Circulating Water Sump, ,

4 it was noted that there were areas in the plant where radioactive j and nonradioactive drains co-exist. I CN 90-13 includes a table to be added to Section 9.0 of the UFSAR which tabulates the exceptions to the statement quoted from

section 9.3.3.2 Item 2. Also the method of control utilized for l compliance with criteria defined in the Standard Review Plan 9.3.3 l Reg Guide 1.143 and IE Cire 80-18 is documented.

I The change to the SAR does not increase the probability of an

! accident previously evaluated because as per section 9.3.3.5

" Safety Evaluation" of the UFSAR the plant drainage systems have no safety function. Accidents described in Section 9.3.3.5 involve exposed drain lines leaking after a seismic event and then l dripping into areas containing essential equipment. The fact that radioactive and non-radioactive drains co-exist in some areas of

the plant, does not alter the probability of existing drain lines leaking after a seismic event. There were no accidents described in Chapter 15 that were related to the proposed SAR Change.

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Therefore, this SAR 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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SAR Change Notice 91-09 This SAR Change Notice describes an evaluation justifying fire load corrections to UFSAR Sections 9.5.1.6.64, 9. 5.1. 6. 25a, 9A. 6.5.1(b) , Table 9A-1 and Table 9A-61.

During a review of the fire Hazards Review Checklist contained in DCP 4KM-0150 (addition of carpet to Hope Creek Control Room area 5510, & 5511 -Fire Area CD 46) it was determined that I rooms the combust 5509,ible fire loading data incorrectly reflected the amount of combustibles added.- This SAR Change Notice provides the correct combustible loading information in the appropriate sections of the UFSAR.

The 2.3 minute fire load increases to 3.3 minutes by the addition of control room area carpets. This is well within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated barriers encompassing.the control room area (fire area CD 46). Therefore, the fire load increase'has no negative effect on fire barrier systems or parameters for this fire area.

Therefore, this SAR 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.

SAR CN 94-40: This change to the SAR updates table 2.2-5

" Estimates of Hazardous Chemical Traffic". The Frequency criteria for river barge traffic as contained in NRC Regulatory-Guide 1.78 is 50 or more trips per year. A comparison of the US Coast Guard j data to the Regulatory Guide 1.78. frequent criteria for river-

! barge traffic concludes that no particular-hazardous chemicals such as those contained in: table C-1 or' Reg Guide 1.78 meets the  ;

" Frequent" criteria. That is, hazardous chemicals such as those '

contained in Table C-1 of Reg Guide 1.78 are shipped past Hope Creek less than 50 times per year.

Since the shipments are considered " infrequent".they need not be considered in the control room habitability analysis. Therefore, further evaluation of a postulated release from a river barge transporting the hazardous chemicals is not required per the guidelines of NRC Regulatory Guide 1.78.

Therefore, this SAR 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.

SAR CN 94-44: This SAR Change Notice is updating the SAR to indicate as-built station conditions. Section 9.5.1.6 "SRP Rule Review" of the SAR describes differences and clarifications to the l Branch Technical Position CMEB 9.5-1. The fact that no local-alarm is provided is not addressed as a clarification in section i 9.5.1.6. A deficiency evaJuation determined that the local alarm j originally was intended to notify the' plant fire brigade of the fire location. Since the control room personnel receive the alarm

( and notify the plant fire brigade of the fire location, the local l alarm is deemed unnecessary.

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f This change will revise the SAR to indicate that no local fire

! alarm is provided. This modification does not effect the  !

I consequences of the fire previously evaluated. This is due to the fact that the fire brigade will be notified and respond, because, a reliable system is in place to notify the fire brigade no matter I where they are on the plant site. Therefore, the modification does not effect the ability of the fire brigade to respond to a fire emergency as previously evaluated.

-fherefore, this SAR 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.

l l SAR CN 94-047 and DEF #DEH-90-00036 : This SAR Change Notice and j DEF Documents discrepancies in HC UFSAR Tables 9A-1 and 9A-2 thru. '

9A-35 for fire areas CD17 thru CD20. There are inconsistencies in these tables concerning combustible loading for cable chases A, B, C, and D in the Auxiliary / Diesel Building at elevations 77' thru 150'. The Equivalent Fire severity will be slightly increased or decreased (in minutes) for the affected fire areas. This Change notice documents the correct combustible loading information in the appropriate sections of the UFSAR. No physical modifications to the facility are required as a result of the change in the fire severity.

This correction does not effect the fire being evaluated. Nor does it effect the fire ignition or the fire involvement area. The equivalent fire severity is a calculation used for comparison i l

only. All areas involved have a three hour rating with fixed fire suppression. Additionally these areas are equipped with automatic fire detection, which summons the plant brigade.

Therefore, this SAR 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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