ML20133M983

Monthly Operating Rept for Dec 1996 for Hope Creek Generating Station

ML20133M983
Person / Time
Site:	Hope Creek
Issue date:	12/31/1996
From:	Harris R, Kepley L Public Service Enterprise Group
To:
Shared Package
ML20133M974	List: ML20133M970 ML20133M983
References
NUDOCS 9701230115
Download: ML20133M983 (10)

Text

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

l r .

h I

l l

l I

l i

INDEX i I

NUMBER  :

SECTION OF PAGES l Operating Data Report . . . .. .. . .. . . . . . ..2 j 1

Average Daily Unit Power Level . . . ..I l Refueling Information.. .. . .. . . .1 l 1

Monthly Operating Summary.. .. . . . .1 i

1 Summary of Changes, Tests, and Experiments. . . . . . . . .4  !

l i

l i )

i r

i i

I t

I i

t l 9701230115 970115 PDR ADOCK 05000354 l R PDR t

I

l l

l l DOCKET NO.: 50-354 l UNIT: Hope Creek j DATE: 01/09/97 l

COMPLETED BY: R. Harris TELEPHONE: (609) 339-1777 OPERATING DATA REPORT OPERATING STATUS

1. Reporting Period December 1996 Gross Hours iu 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) Ng.ne n

4. Reasons for restriction (if any)

This Month Yr To Date Cumulative

5. No. of hours reactor was critical 744.0 6799.2 73723.1 i 6. Reactor reserve shutdown hours 0.0 0.0 0.0

7. Hours generator on line 744.0 6618.4 72560.0 1

8. Unit reserve shutdown hours 0.0 0.0 0.0

9. Gross thermal energy generated (MWH) 2418776 21093998 231868247

10. Gross electrical energy generated (MWH) 822480 7068591 76894213

11. Net electrical energy generated (MWH) 790643 6754484 73471719

12. Reactor service factor 100.0 77.4 83.8

13. Reactor availability factor 100.0 77.4 83.8

14. Unit service factor 100.0 75.3 82.5

15. Unit availability factor 100.0 75.3 82.5

16. Unit capacity factor (using MDC) 103.1 74.6 81.0

17. Unit capacity factor (using Design Mwe) 99.6 72.1 78.3

18. Unit forced outage rate 0.0 0.0 4.6

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

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

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

. l' .

DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 01/09/97 COMPLETED BY: B. Harris TELEPilONE: (609) 339-1777 OPERATING DATA REPORT

! UNIT SHUTDOWNS AND POWER REDUCTIONS MONTH DECEMBER 1996 METHOD OF SHUITING DOWN THE TYPE REACTOR OR i F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (H%RS) (1) POWER (2) ACTION / COMMENTS n/a b

I l

, 1-DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 01/09/97 COMPLETED BY: R. Harris TELEPHONE: (6091 339-1777 AVERAGE DAILY UNIT POWER LEVEL '

MONTH PECEMBER 1996 DAY AVERAGE DAILY POWER LEVEL DAY. AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 1064 17 1069 2 1061 18 1042 3 1063 19 101.5 4 1063 20 1069 ,

5 1062 21 1065 6 1065 22 1066 7 1062 23 1073 8 1050 24 1049 9 1062 25 1066 10 1070 26 1067 11 1066 27 1072 12 1055 28 1063 13 1061 29 1065 14 1056 30 1048 15 1063 31 1064 16 1055 l

4

DOCKET NO.: 50-354 UNIT: Hope Creek DATE: 01/09/97 COMPLETED BY: L. Keplev TELEPHONE: (609) 339-1106 REFUELING INFORMATION MONTH DECEMBER 1996

1. Refueling information has changed from last r onth:

Yes No X

2. Scheduled date for next refueling (RF07): 9/6/97

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)?

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

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

Not requig_d_

6. Important licensing considerations associated with refueling:

N/A

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage 1472

8. Present licensed spent fuel storage capacity: 4006 Future spent fuel storage capacity: 4006

9. Date oflast refueling that can be discharged 5/3/2006 to spent fuel pool assuming the present licensed capacity: (EOCl3)

(Does allow for full-core off-load)

(Assumes 244 bundle reloads every 18 months until then) l (Does no_1 allow for smaller reloads due to improved fuel) 1 l

. l l

l DOCKET NO.: 50,354 UNIT: Hope Creek j DATE: 01/09/97 l COMPLETED BY: R. Harris TELEPHONE: (609) 339-1777 l 1

MONTHLY OPERATING

SUMMARY

l MONTH DECEMBER 1996 i The Hope Creek Generating Station remained on-line for the entire month and operated at essentially 100% power for the month of December 1996 except one load reduction identified below. l Power was reduced to 87% on December 8,1996 starting at 0234 hours0.00271 days <br />0.065 hours <br />3.869048e-4 weeks <br />8.9037e-5 months <br /> to perform .

monthly turbine valve testing. The unit was returned to 100% power on December 8, {

1996 at 1048 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.98764e-4 months <br />.

At the end of the month the unit had been on-line for 54 days.

l l

1 r

i

, i

I.

l l DOCKET NO.: 50-354 l UNIT: Hope Creek l DATE: 01/09/97 l COMPLETED BY: L. Keolev TELEPHONE: (609)339-1106

~

SUMMARY

OF CHANG l STS, AND EXPERIMENTS FOR THE HOPE CREEA GENERATING STATION i l MONTH DECEMBER 1996 l l

The following items completed during November 199&ve been evaluated to determine:

1. If the probability of occurrence or the consequences of an accident or malfunction of I equipment important to safety previously evaluated in the safety analysis report may be increased, or l

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 i to the plant nor did they affect the safe shutdown of the reactor. These items did not change the '

plant efiluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmenti questions are involved.

Design Changes Summary of Safety Evaluations 4EC -03471 Package 1, InstallF.iberO pti,M uliPairVobe and D ata C abling. This design change provides increased telecommunications capability for the Hope Creek Auxiliary and Radwaste Buildings, and the Computer Equipment Rooms. The cab!es installed to support the design have no safety-related function and do not interface with plant operating systems. The conduits and trays are supported as Seismic Category II/I.

The telecommunication racks being added are either Seismically II/I mounted or their failure in a seismic event cannot cause damage to safety related ec.uipment. The increase in combustible loading due to the replacement cables is within the design basis for the affected areas.

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

4E C -03599 Package 1, Servbe W ater Travelng Screen Diffenmtil Level M anioring System Upgrade This change replaces the Hope Creek Service Water Traveling Water Screen ultrasonic differential level monitoring system with a bubbler type differential level monitoring system. The modified differential water level control system will still automatically actuates the traveling water screen to high speed upon detection of l a differential level condition across the screen which exceeds 10 inches. The design of the l bubbler differential level monitoring system incorporates a fail safe design feature (Class IE flow switches) and ensures that the Class lE portions of the bubbler channels for each traveling water screen are isolated and physically separated from each other to satisfy the single failure criterion. This design ensures that the redundancy afforded by the design of i the Station Service Water System (SSWS) is maintained. Should a failure of the non-l

1 Class lE >ortion of the bibbler system occur, the traveling screens will be actuated to high speec due to the fail safe design feature afforded by the Class lE flow switches. This fail safe design does not impact the traveling water screen components, since they have been designed for continuous operation at high speed. The Class lE equipment will be installed m accordance with Seismic Category I requirements, and the non-Class lE equipment will be installed in accordance with the requirements of Seismic Category II/I.

The proposed modification will not increase the probability of a missile event, fire, or flood previously evaluated in the SAR. The change will not affect the ability of the SSWS to perform its safety related function.

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 i Question.

4H E -00356 Package 4, R em ove Fbw Switch 1EPFS-2225C and InstallG asket and j B lind Fhnge to Restore Integr1y of Pipe Line E P-6"-H ZC -002. This design change removes the service water screen and backwash flow switches (IEPFS-2225 A, B, C, D) and installs gaskets and blind flanges in place of the flow switches. Stress and seismic evaluations have been performed to demonstrate the adequacy of the change. The replacement of the flow switches with blind flanges does not affect the amount of flow being provided to the Reactor Auxiliary Cooling System (RACS) and Safety Auxiliary Cooling System (SACS) heat exchangers during 31 ant conditions, transients, and accident conditions. The change does not: 1) change, c egrade, or prevent actions described or assumed in any accident described in the SAR, 2) alter any assumptions previously 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 S AR,4) : iect a fission product barrier, or 5) change the composition or inventory of radioactivity releases. This change does not adversely affect the operability of the Station Service Water System (SSWS).

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.

Procedures Summary of Safety Evaluations H C D P-SO E A -0001 Q ) Revisbn 12, Station Servjoe W atEr System GSW S) o peratbn. The revised SSWS operating procedure, HC.OP-SO.EA-0001(Q), provides a method to refill an out-of-service SSWS loop by diverting flow from the in-service loop through the Reactor Auxiliary Cooling System (RACS) cross-tie line. System Calculation EA-0032, SSWS Loop Fill Using RACS Cross-tie, was performed to evaluate the effects of filling an inoperable SSWS loop from the operable loop. The calculation provided the following conclusions:

With the ultimate heat sink (UHS) temperature less than 74.9oF, the RACS cross-tie valve can be fully opened without impacting the capability of Safety Auxiliary Cooling System (SACS) to cool its safety related heat loads and without subjecting the SSWS pumps to runout conditions.

With the UHS temperature between 74.9aF and 80oF, the out-of-service SSWS loop can be filled at a rate such that indicated flow of the operable loop increases by not greater than 2124 gallons per minute (GPM). Since a +/- 300 GPM for instrument inaccuracies was assumed in the calculation, the procedural limit will be

i set at a maximum of 1800 GPM. Under this condition, the operabie SSWS loop retains the capability to remove safety-related heat loads under accident conditions.

By incorporating the preceding limite into HC.OP-SO.EA-0001(Q), the capability of the operable SSWS loop is neither challei:ged nor degraded when filling an out-of-sersice SSWS loop. The change does not impact the capability of the operable SSWS loop to perform its intended safety function and will not pose any operational challenges to the SSWS pumps. This change will ensure adequate cooling to systems required to mitigate l the consequences of an accident and will ensure that plant configuration remains within full compliance with Hope Creek Technical Specifications while restoring a drained SSWS  ;

loop to service. The change does not afTect or introduce any new accident precursors, nor l does it introduce physical change to plant systems, structures or components. l 1

Therefore, implementation of this procedure revision does not increase the probability or conseguences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.

N C N A -A P 2Z -0001 O ) Revisbn 9, Nuchar Businem Unit NBU) Procedure ,

Syman The proposed changes are administrative in nature and integrate present policy l procedures into other documents and establish a four tier procedure hierarchy. The l proposed changes will move pertinent information into other documents that are more useable and are reviewed more frequently by a larger population of NBU employees, thereby, integrating policy information more frequently into daily NBU activities. In addition, this revision proposes a change to clarify section 13.5 of the Hope Creek UFSAR to differentiate between On-The-Spot-Change and a partial procedure. The  ;

proposed changes have no direct effect on any systems, structures or components described in the UFSAR. The changes do not propose any reduction in the qualifications and training requirements of personnel, the management control and authority, administrative program commitments or quality assurance program. This change meets the intent of the Standard Review Plan, NUREG 0800. Therefore, the proposed changes do not reduce the margin of safety, result in a condition not previously evaluated and do not change the way the quality assurance program is implemented. The revised organization should improve alignment of responsibilities and management focus on activities related to plant safety.

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

Other Summary of Safety Evaluation Safety Evahatbn H EA -M SE-0864, R ev. O, Post W eH H eat Treatm ent of W eE Repaim forSA A 351,G rade CD4M Cu. An NRC Safety Evaluation Report approving weld repairs to CD4MCu material used in the Service Water Pumps required post weld heat treatment (PWHT) after weld repairs. Although PWHT was performed following weld repairs of pressure boundary components, PWHT of the impeller (a non pressure boundary component) following weld repair was not performed. This safety evaluation justified use ofimpellers made of CD4MCu material that had been weld repair and not PWHT. The overall wear / corrosion resistance of the CD4MCu material is better than the current nickel aluminum bronze impellers. The pump vendor has certified that the CD4MCu components when installed in the Service Water Pumps will meet performance and design requirements of the ASME code and the original pumps. No new failure

~

l l

i modes have been introduced by the CD4MCu impellers without PWHT as the failure mode is the same as for the present pump material. The new pumps are certified and tested to meet ASME III requirements.

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

Temporary Modifications Summary of Safety Evaluations l

Deficiency Reports Summary of Safety Evaluations l UFSAR Change Notices Summary of Safety Evaluations There were no changes in these categories implemented during November 1996.

Correction of November Monthly Operatine Report, dated December 13,1996 Design Change (DCP) 4HM-0258-2, Permanent Lifling Points, and Temporary Modification 96-025, Battery IBD411 Ceu No.19 Jumpered Out were incorrectly reported as being installed during the month of October. DCP 4HM-0258-2 and Temporary Modification 96-025 were not installed at Hope Creek Generating Station during the month of October.

l I

4 l