ML20137X731

From kanterella
Jump to navigation Jump to search
Monthly Operating Rept for Mar 1997 for Hope Creek
ML20137X731
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 03/31/1997
From: Phillips R
Public Service Enterprise Group
To:
Shared Package
ML20137X730 List:
References
NUDOCS 9704220148
Download: ML20137X731 (9)
v • d • e


Text

_ _ _ _ . m _ _ . _ . . . . _ . _ . . _ . _ _ _ . . - . . _ . _ . . . _ _ ..- - __ .. _ _ _. _ . . - . . _ _ _

DOCKET NO.: 50-354 y UNIT: Hone Creek l_

DATE: 04/08/97 COMPLETED BY: R. Phillins TELEPHONE: (609) 339-2735 l_ OrMRATING DATA REPORT l OPERATING STATUS

1. Reporting Period March 1997 Gross Hours in Report Period 211
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 Cumulativ Month Date a

5. No. of hours reactor was 744.0 2160 75883.1 critical
6. Reactor reserve shutdown RA2 222 Am2 hours

'7 . Hours generator on line lii 2160 74720

8. Unit reserve shutdown hours azQ HzQ 92Q
9. Gross thermal energy 2411356 7004237 238872484 generated (MWH)
10. Gross electrical energy 818720 2379970 79274183 l generated (MWH)
11. Net electrical energy 787666 2289595 75761314 generated (MWH)
12. Reactor service factor 100.0 100.0 84.2
13. Reactor availability factor 100.0 100.0 84.2
14. Unit service factor 100.0 100.0 82.9
15. Unit availability factor 100.0 100.0 82.9
16. Unit capacity factor (using 102.7 102.8 81.5 MDC)
17. Unit capacity factor (using 99.2 992 1 78.8 Design MWe)
18. Unit forced outage rate 222 gig Azh
19. Shutdowns scheduled over next 6 months (type, date, &

duration):

l Refueling Outage, September 6, 1997, 60 days

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

9704220148 970414 PDR ADOCK 05000354 R PDR i _

DOCKET NO.: 50-354 l

UNIT: Hone Creek DATE: 04/08/97 COMPLETED BY: R. Phillins TELEPHONE: (609) 339-2735 OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS I

NONTH H&&CH 1997 i METHOD OF

! SHUTTING DOWN THE TYPE REACTOR OR

F= FORCED DURATION REASON REDUCING CORRECTIVE l

NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS l n/a I

i l

I i

l l

l DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 04/08/97 COMPLETED BY: R. Phillios ,

TELEPHONE: (609) 339-2735 AVERAGE DAILY UNIT POWER LEVEL i MONTH MARCH 1997 1

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) l 1 995 17 1066 2 1049 18 1062 3 1067 19 1965 4 1061 20 1061 5 1067 21 1059 l

6 1057 22 1051 7 1064 23 1094 8 1058 24 1039 9 1068 25 1078 10 1068 26 1051 l

l 11 1052 27 1055 12 1063 28 1051 13 1067 29 1053 14 1025 30 1020 15 1099 31 1063 16 1065 l

l l

l r

i

l .

DOCKET NO.: 50-354 i

UNIT: Hooe Creek

( DATE: 04 /08]J2 l COMPLETED BY: L. Keolev l TELEPHONE: (609) 339-1106

! REFUELING INFORMATION MONTH _MARCE_J,997 i

1. Refueling information has changed from last month:

Yes __ No X

2. Scheduled date for next refueling (3F07): 9/6/92
3. Scheduled date for restart following refueling:

11/5/97 4A. 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 (SORC)? i Yes __

No X If no, when is it scheduled? To Be Determined for Cycle 8 COLP

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

Mgt recuired.

6. Important licensing considerations associated with refueling:

EIA

7. Number of Fuel Assemblit:3:

A. Incore 764 B. In Spent Fuel Storage 1472

8. Present licensed spent fuel storage capacity:4006 Future spent fuel storage capacity: 1006 9.- Date of last refueling that can be dischargedS/3/2006 to spent fuel pool assuming the present licensed capacity: l (EOCl3) i l 1 (Does allow for full-core off-load)

(Assumes 244 bundle reloads every 18 months until then)

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

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

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

DOCKET NO.: 50-354 .

UNIT: Hone Creek DATE: 04/08/97

, COMPLETED BY: R. Phillios l TELEPHONE: .(609) 339-2735 i

j l MONTHLY OPERATING

SUMMARY

l l MONTE MARCH 1997 ,

l

  • The Hope Creek Generating Station remained on-line for the entire month and operated at 100% power for the month L of March 1997. There were two load reductions which are '

identified below.

  • Power was reduced to'87% on March 1, 1997, starting at l 0105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br /> to perform monthly turbine valve testing. The unit was returned to 100% power on March 1, 1997, at 1325 i hours.
  • Power was reduced to 87% on March 30, 1997, starting at 0209 hours0.00242 days <br />0.0581 hours <br />3.455688e-4 weeks <br />7.95245e-5 months <br /> to perform turbine valve testing and rod shuffle. The unit was returned to 100% power on March 30, 1997, at 1045 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.976225e-4 months <br />.
  • At the end of the month the unit had been on-line for 144 l days.

l I

i i

?

r i

, 4 I

l.

1 1

DOCKET NO.: 50-354 UNIT: HoDe Creek DATE: 04/08/97 COMPLETED BY: L. Keolev

) TELEPHONE: (609) 339-1106

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEE GENERATING STATION MONTH MARCH 1997 The following items completed during February 1997 have been evaluated to determine:

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 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 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 or environmental questions are involved.

Desion Chances Summary of Safety Evaluations Replacement of Air-Operated Valves in the Safety Auxiliary Cooling System (SACS), 4EC-03612 Pkgs. 2, 4, 6, and 7. The Standby Diesel Generator (SDG) room cooler water supply valves in the SACS were replaced. The previous valves were carbon steel air actuated flexible-wedge gate valves. The replacement valves are stainless steel air actuated ball valves which are more suitable to the application. UFSAR Figure 9.2-5 was changed to show the type and material of the replacement valves and actuators. UFSAR Table 3.9-18 was changed to show the valve / actuator replacement type.

This replacement does not change the function of the valves or the function of the affected a,ystems. The change does not alter the ability of the SACS, the SDG Room Recirculation System, or the SDGs to perform their intended

. safety function. The affected systems will function in

~

accordance with the original design and licensing basis of the plant.

I

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

1

{ .

i

Therefore, this design change does not increase the
- probability or consequences of an accident previously j described in the UFSAR and does not involve an Unreviewed l Safety Question.

i

! Replacement of Air-Operated Valves in the safety Auxiliaries l Cooling system (SACS), 4EC-03612, Pkg. 9. The Residual Heat i

Removal (RHR) pump room cooler water supply valves in the i SACS were replaced. The previous valves were carbon steel ,

air actuated flexible-wedge gate valves. The replacement i valves are stainless steel air actuated ball valves which i
are more suitable to the application. UFSAR Figure 9.2-4 l was changed to show the type and material of the replacement valves / actuators. UFSAR Table 3.9-18 was changed to show j the valve / actuator replacement type. This replacement does  !

j- not change the function of the valves or the function-of the affected systems. The change does not alter the ability of i the' SACS, the Equipment Area Cooling System (EACS), or the '

RHR pump room unit coolers to perform their intended safety J functions. The affected systems will function in accordance  :

with the original design and licensing basis of the plant.

Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed ,

Safety Question.

Replacement of Air-Operated valves in the safety Auxiliaries i Cooling system (SACS), 4EC-03612, Pkg. 27. The Core Spray -

(CS) pump room cooler water supply valves in the SACS were replaced. The previous valves were carbon steel air I actuated flexible-wedge gate valves. The replacement valves 1 are stainless steel air actuated ball valves which are more suitable to the application. UFSAR Figure 9.2-4 was changed to show the type and material of the replacement valves / actuators. UFSAR Table 3.9-18 was changed to show- -

the valve / actuator replacement type. This replacement does l not change the function of the valves or the function of the affected systems. The change does not alter the ability of the SACS, the Equipment Area Cooling System (EACS), or the CS pump room unit coolers to perform their intended safety functions. The affected systems will function in accordance with the original design and licensing baris of the plant.

Therefore, this design change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.

Reacve Flow switch 1EPFs-2225B, 4HE-00356, Pkg. 2.. This design change removed the service water screen and backwash

l \

) ..

I

. -1 i

f . .

l i flow switch, 1EPFS-2225B, and installed gaskets and a blind j flenge in place of the flow switch. Stress and seismic evaluations have been performed to demonstrate the adequacy of the change. The replacement of.the flow switch with a j blind flange does not affect the amount of flow being

provided-to the Reactor Auxille.ry Cooling System (RACS) and I Safety Auxiliary Cooling System (SACS) heat exchangers i during plant conditions, transients, and accident

' conditions. The change does not: 1) change, degrade, or l prevent actions described or assumed in any accident

described in the SAR, 2) alter any assumptions previously I

i made in evaluating radiological consequences of any accident described in the SAR, 3) affect the mitigation of the radiological consequences of any accident described in the

! SAR, 4) affect a fission product barrier, or 5) change the i composition or inventory of radioactivity releases. This

change does not adversely affect the operabilb 7f the Station Service Water System (SSWS).

i Therefore, this design change does not increase the j probability or consequences of an accident previously i

i described in the UFSAR and does not involve an Unreviewed Safety Question.

j Procedures 8"--=ry of Safety Evaluations l

TEC.CE-SA.RC-0004(Q), Temporary Balance of Plant (BOP)

{~ sampling Program. This temporary BOP Sampling Program )

establishes chemistry parameters, specifications, sampling frequencies, and; sampling locations to meet Updated Final q

Safety Analysis Report (UFSAR) sampling requirements during i

! the BOP sample station replacement. The proposed procedure j

l. does.not have any negative effects on safety related '

functions and does not compromise any safety related system g or components or prevent safe shutdown of the plant.

[

Therefore, this temporary procedure does not' increase the

. probability or consequences of an accident previously

^

described in the UFSAR and does not involve an Unreviewed Safety Question.

i i Other sa===ry of safety Evaluation j safety Evaluation B97-006, On-Site Transportation of Sales j Unit l's Original Steam Generators (OSGs) including

- Temporary Parking of the 08Gs. A temporary parking area, l for storage of the OSGs prior to their shipment offsite, was j . erected near the low level waste storage facility, by

{ creating a Radiological Controlled Area (RCA) and erecting temporary concrete block shielding. This safety evaluation i was prepared-to evaluate the effect of the RCA on the Hope 4

4

  • Creek Generating Station. The safety evaluation demonstrated that neither the RCA, nor the transportation, nor the temporary parking of the OSGs adversely affected any structure, system or component. The temporary parking area, l

l as erected and controlled, does not present any radiological safety concerns.  ;

! Therefore, implementation of this change does not increase the probability or consequences of an accident previously

! described in the UFSAR and does not involve an Unreviewed Safety Question.

UF8AR Chance Notices Sn==?ry of Safety Evaluations Temporary Modifications Sn==ery of Safety Evaluations

(

Deficiency Reports Sn==?ry of Safety Evaluations I

There were no changes in these categories implemented during l February 1997.

i l J l

t l

t l

2 i

Retrieved from "https://kanterella.com/w/index.php?title=ML20137X731&oldid=2827813"