ML20041G201

Monthly Operating Rept for Jan 1982

ML20041G201
Person / Time
Site:	Davis Besse
Issue date:	02/10/1982
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML20041G196	List: ML20041G195 ML20041G201
References
NUDOCS 8203190387
Download: ML20041G201 (14)

Text

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AVERAGE DAILY UNIT POWER LEVEL 0

DOCKET NO. _50-346 UNIT Davis-Besse Unit 1 DATE February 10, 1982 COMPLETED BY _Bilal Sarsour TELEPflONE _419-259-5000. Ext.

384 MONTH danuary, 1982 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net)

(MWe-Net!

790 37 878-18 882 3

798 19 883 4

861 20 885 5

883 21 856 6

884 22 729 7

883 23 571 8

881 24 573 9

886 25 574 10 884 26 574 11 882 27 574

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12 882 28 573 13 862 29 574 l

14 875 573 39 i

is 883 572 l

3 16 773 INS'iltUCTIONS On this (orinat.hs the aver.sge daily unit power level in MWe. Net for cach day in the reporting inonth the nearest whole nwgawaii.

(9/77) l l

P 8203190387 820210 PDR ADOCK 05000346 R

PDR

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OPERATING DATA REPORT Y

DOCKET NO.

50_-346 DATE 2/10/82 COMPLETED llY R41n1 Sarsour TELEPilONE 419-N_-5000, Ext.

OPERATING STATUS 384

1. Unit Name:

Davis-Besse Unit 1 N0t'5

2. Reporting Period:

January, 1982

3. Licensed Thermpi Power (MWt):

2772

4. Nameplate Rating (Gross MWe):

925 S. Design Electrical Rating (Net MWe):

906

6. Maximum Dependable Capacity (Gross MWe):

918

7. Maximum Dependable Capacity (Net MWe):

874

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

The maximum dependable capacity now reflects the summer rating obtained during the most restrictive seasonal conditions.

9. Power Level To Which Restricted,if Any (Net MWe):

10. Reasons For Restrictions.lf Any:

This Month Yr.-to Date Cumulative

11. Ilours in Reporting Period 744 744 30,745

12. Number Of flours Reactor Was Critical 744 744 16.974

13. Reactor Reserve Shutdown flours 0

0 3,334.7

14. Ilours Generator On Line 744 744 15,994.2

15. Unit Reserve Shutdown llours 0

0 1,731.4

16. Gross Thermat Energy Generated (MWil) 1.822.202 1.822.202 36.943.727

17. Gross Electrical Energy Generated (MWil) 608,849___,

608,849

_12,291,100

18. Net Electrical Energy Generated (MWil) 575,941 575,941

_11,473,226

19. Unit Service Factor 100 100 52.0

20. Unit Availability Factor 100 100 57.7

21. Unit Capacity Factor (Usmg MDC Net) 88.6

22. Unit Capacity Factor (Using DER Net) 85.4 _

88.6 42.7 85.4 41.2

23. Unit Forced Outage Rate 0

0 24.0

24. Shutdowns Scheduled Over Next 6 Months (Type. Date.and Duration of Each t:

Refuelinn Outane 2/26/82 - 5/21/82

25. If Shut Down At 1.nd Of itepost Period. Es'imated Date of Startup:

26. Units in Test Status irrior io Commercial Operation):

Fusecast Achiesed INITIAL CRITICA LITY INITI A L ELECTRICITY COMMERCIA L OPER A TlON (9/77 )

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50-346 UNIT SHUTDOWNS AND POWC: REDUCTIONS DOCKET NO.

UNIT NAME Davis-Besse Unit ~1 t

DATE February 10. 1982 COMPLETED BY Bilal Sarsour Il REPORT MONTH January, 1982 TELEPIIONE 419-259-5000. Ext. 384 l

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j.Y 5 Licensee Eg ja Cause & Correctise l

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No.

Date g

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,R 1g5 Event g,7 E-3 Action to

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Present Recurrence-P l

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1, 82 01 22 F

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5 Power was reduced to 687,due to a leaking extraction steam line expan-sion bellow in No..I high pressure condenser to No. 4 high pressure heater.

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4 F: Forced Reason:

Method:

Exhibit G Instructions S: Schedu!cd A Equipment Failure (Explain) 14tanual for Preparation of Data B41aintenance of Test 241anual Scrani.

Entry Sheets for Licensee C-Refueling 3 Automatic Scram.

Event Report (LER) File (NUREG-D RcEulatory Restriction 4 Continuation from Previous Month 0161)

E-Operator Training & License Examination 5-Load Reduction F-Admimstratise 9-Other (Explain) 5 G Operational Eiror (Explain)

Estiibit I Sanw Soaree i

(9/77)

Il-Ot her ( E splain)

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OPERATIONAL

SUMMARY

January, 1982 i

t 1/1/82 - 1/31/82 Reactor power was maintained at 89% with the generator gross load at approximately 830 1 10 MWe until JS05 hours on January 4, 1982. Reactor power was increased and attained

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100% full power on January 5, 1982, i

j On January 13, 1982, while the unit was at approximately

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100%, power was reduced to approximately 90% due to Borated j

Water Storage Tank (BWST) temperature being outside the Engineering Safety Features (ESF) analysis of the Final Safety j

Analysis Report (FSAR).

The reactor power was slowly increased and attained 100% full power on January 14, 1982. On January 16, 1982 while the unit l

was at 100%, power was reduced to approximately 50% for axial power shaping rod (APSR) withdrawal. Reactor power was. then 1

increased and 100% full power was achieved by January 17, 1982.

Power was maintained at 100% until 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br /> on January 21, 1982 when power was reduced to approximately 88% due to a leak-ing extraction steam line expansion bellow in No. 1 high pres-f sure condenser to No. 4 high pressure heater.

2 On January 22, 1982, while the unit was at approximately 88%

power, power was reduced to 68% to prevent further degradation of expansion joints.

Reactor power was maintained at 68% for the remainder-of the month.

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REFUELING INFORMATION DATE:

January, 1982 1.

Name of facility:

Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown:

February 26, 1982 3.

Scheduled date for restart following refueling:

May 21, 1982 4.

Will refueling or resumption of operation thereafter require a technical specifi-cation change or other license amendment? If answer is yes, what in general, will these be?

If answer is no, has the reload fuel design and core configura-tion be'n reviewed by your Plant Saiety Review Committee to determine whether e

any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?

Final reload analysis is scheduled for completion in March, 1982.

No technical spec'ification changes or other license amendments identified to date.

5.

Scheduled date(s) for submitting proposed licensing action and supporting infor-mation.

February, 1982 6.

Important licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or performance analynin methods, significant changes in fuel design, now operating procedures.

None identified to date.

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7.

The number of fuci assemblics (a) in the core and (b) in the npent fuel nturane I

pool.

44 - Spent Fuct Annemblien (a) 177 (b) 48 - New I'uci Assemblies 8.

The present licensed spent fuel pool storage capacity and the size of any in-crease in licensed storage capacity that has been requested or is planned, in number of fuel assemblics.

Present 7 15 Increanc size by 0 (zero) 9.

The projected date of the laut refueling that can be dincharned to tho : pent fuel pool annoming the preuent licenned capacity.

Date 1988 - nasuming abilLty to unload the entire core into the spent fuci pool is maintained.

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COMPLETED FACILITY CHANCE REQUESTS FCR NO:

77-065 SYSTEM:

Post Accident Containment Radiation Monitoring System COMPONENT: Radiation Monitors CHANGE, TEST OR EXPERIMENT:

Facility Change Request 77-065 was implemented to disconnect the purge initiation on the following radiation monitors:

RE 1003A RE 5327 RE 1003B RE 5328 RE 2024 RE 5403 RE 2025 RE 5405 RE 5029 RE 8432 RE 5030 RE 8433 RE 5052 RE 8434 RE 8442 The work was completed April 22, 1981.

REASON FOR CHANGE:

These monitors have no purge piping.

It is desired to disconnect the purge push buttons to prevent an inadvertent release of radioactivity.

SAFETY EVALUATION:

The work for FCR 77-065 involved disconnecting the purge buttons for the previously mentioned monitors. Of these monitors, only RE 5327, RE 5328, RE 5029 and RE 5030 are Q-listed. The disconnecting of the purge buttons does not affect the safety function of these monitors and therefore, no unreviewed safety question exists.

COMPLETED FACILITY CilANGE REQUESTS FCR NO:

78-450 SYSTEM: Makeup & Purification (Boron Injection Flowpath)

COMPONENT:

Boron Injection Flowpath ifeat Trace CilANGE, TEST OR EXPERIMENT:

Facility Change Request 78-450 was issued to relocate the thermo-couple cn circuit 153 of the Boron Injection Flowpath lieat Tracing approximately three (3) feet away from the "T" of the letdown line. The work for this FCR was completed May 21, 1980.

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REASON FOR CIIANGE:

The thermocouple on circuit 153 of the Boron Injection Flowr'th lleat Tracing was located approximately one foot away from the "T" of the letdown line.

At full power, the letdown temperature is approximately 100 F.

The colder temperatures in the letdown line affected the thermo-couple readings. The old thermocouple location did not reflect the actual temperature in the entire section of the heat traced portion of the pipe.

SAFETY EVALUATION:

The Tech. Spec. Weekly Surveillance Test requirements, Section 4.1.2.2 require that the pipe temperature of the heat traced portion of the flowpath of the boron inject. ion should be 2 105 F.

The thermocouple was located approximately one foot away from the "T" and read around 100 F.

The station took temperature readings at three separate points varying from 2 to 3 feet away from the "T" (joint between the letdown line and the boron injection path) using a contact pyrometer and the readings were consistently found to be approximately 115 F This FCR calls for moving the thermocouple to about three feet away f rom the "T" j oint.

Therefore, the change per this FCR does not affect the operability of the boron injection flow path.

In summary the above change does not involve an unreviewed safety question because:

1.

The probabili ty of occurrence or t he consequence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR has-not been increased.

2.

The possibility of an accident or malfunction of a different type t.ha t is not bounded by the previous analysis in the FSAR has not been created.

3.

The margin of safety as defined in the basis for any Davis-Besse Unit 1 Technical Specification has not been decreased.

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e COMPLETED FACILITY CHANGE REQUESTS FCR NO:

78-540 SYSTEM: 416 KV Switchgear "CD" t

COMPONENT: Computer Alarm Q428 g CHANGE, TEST OR EXPERIMENT:

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FCd'78-540 changed computer Alarro Q428, Essential Transfer Bus CD, to bypass ACD2 and ACD3 when AC108 or AD108 is racked 'out.

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shown below:

This change is

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52H/AC108 52H/AD108'

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29 52S 29 39 ACD2 ACD3 ACD4 ACD5 This arrangement will alarm if ACD2, 3, 4 & 5 are not closed.

The work for this FCR was completed May 31, 1980.

t REASON FOR CHANGE

Alarm Q428 is in the alarm state at 'all t,imes because loss of control power in cells ACD2 and ACD3 when AC108 or )\\D108 are racked out.

SAFETY EVALUATION:

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,e FCH 78-540 involved the installation.

alarm Q428 as described above.

f raceway to change computer The instr.11ation in accordance with the Q core' drill report wil.1 assure that no new adverse environments are created.

An unreviewed safety question does not exist.

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J COMPIITED FACILITY CHANGE REQUESTS FCR NO:

79-077 SYSTEM; Auxiliary Feedwater System COMPONENT: HV106, HV360, EV3870 CHANGE, TEST OR EXPERIMENT:

Under Facility Change Request 79-077 valves HV106, ifV360 and HV3870 were modified by installing direct current powered motor operators. The work was completed August 22, 1980.

_ REASON FOR CHANGE:

This was a license condition requirement (2.c(3).(f)) to be completed during the first refueling outage.

SAFETY EVALUATION:

Final Safety Analysis Report (FSAR) Section 9.2.7.3 is the safety evalua-tion. Per Section 9.2.7.31 of the FSAR, " System reliability is achieved by the following features:

DC power will be supplied to the auxiliary feedpump steam inlet valve MS-106 and pump outlet valves AF3870 and AF360 (of one train) to ensure diverse electric power sources to the valves of the auxiliary feedwater sys-tem required'to open for system functioning."

COMPLETED FACILITY CHANGE REQUESTS FCR NO:

79-105 SYSTEM:

Clean Liquid Radwaste COMPONENT: Clean Waste Monitor Tanks CHANGE, TEST OR EXPERIMENT:

Facility Change Request 79-105 was implemented to revise the control and alarm circuits for the Clean Waste Monitor Pumps and Tank and computer alarm Q141, Clean Waste Monitor Tank and Filter. The changes made were as follows:

1)

The pump motors were interlocked with the low level alarm so that if the pumps are off, no alarm signal is given.

2)

The pump motors were interlocked with the low flow alarm so that there is no flow alarm if the pumps are off.

REASON FOR CHANGE:

These changes were made to eliminate nuisance alarm Q141.

SAFETY EVALUATION:

Portions of the work package required for implementation of this FCR are nuclear safety related due to a "Q" core drill.

Installation of the core drill in accordance with the core drill cutout report will ensure that a new adverse environment will not be created. An unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO:

79-137 SYSTEM: Essential Valves COMPONENT:

Local PB Stations Cl!ANGE TEST OR EXPERIMENT:

Facility Change Request 79-137 provided locks on handwheels and on local push button stations so that the valves listed on Attachment I cannot be inadvertently moved or operated at the valve or outside the control room.

REASON FOR CllANGE:

This FCR was implemented to prevent the inadvertent position changing of the various valves listed on Attachment 1.

These valves should not have their positions changed, to prevent the defeat or degradation of a safety system, damage of equipment, or an occurrence of an unanalyzed condition on a valve position.

SAFETY EVALUATION:

Locking handwheels on these motor operated valves will prevent the inadvertent positioning of these valves.

Position changes on these valves could defeat or degrade the safety function of the associated systems, damage equipment, or result in unanalyzed conditions. Therefore, the locking will enhance nuclear safety.

The locking of the handwheel mechanism will not compromise the seismic qualfication of these valves, because of the small amount of added weight.

This FCR also includes the design of the locking mechanism for the local push button stations.

This locking mechanism will be seismically installed and no new adverse conditions will result.

This locking of the local push button sLations will prevent iriadvertent mispcsit ioning of these valves. A misposition of these valves could defeat or degrade the safety functions of the associated safety systems, damage the equipment or result in unanalyzed conditions.

The safety of the unit will be enhanced by locking these push button stations.

This is not a ri unreviewed safety question.

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i FCR 79-137 ATTACliMENT 1 If the position is changed from normal on the following valves, a i

t safety system could be defeated or degraded and -afety system equ pmen could be damaged:

Norm.

Safe Pos.

Valve No.

Description Open AFP 1 Suct. Vlv from Cnds Tk.

Open IIV-786 AFP 2 Suet. V1v from Cnds Tk.

Open llV-790 DH Pmp 1 Suct. Vlv from BWST Open IIV-2733 DH Pmp 2 Suct. V1v from BWST Closed llV-2734 Ctmt Emer Sump Viv Closed llV-Dil9A Ctmt Emer Sump Vlv Open liv-D119B IIV-DlI7A BWST Outlet Vlv Open BWST Outlet V1v Open IIV-Dil7B HPI Pmp 2 Recirc Vlv Open IIV-IIP 31 VII-llP32 IIPI Pmp 2 Recirc Viv If the position is changed from normal on the following valves, a safety system could be defeated or its performance degraded:

Norm. Pos.

Valve No.

Description Open SG 2 AFW Iso Vlv Open IIV-599 SG 1 AFW Iso V1v Open llV-608 LP Inj 2 Iso Vlv Open Dif-1A Dli-1B LP Inj 2 Iso Vlv Closed Vent Closed llV-CF5A CFT 2 N2 Vent Closed IIV-CFSB CFT 1 N2 llV-CF2A CFT 2 Liquid Drain Closed CFT 1 Liquid brain llV-CF2B if the position i:, changed from nor :1 on the following valves, safety system equipment could be damaged:

Norm. Pos.

Valve No.

Description Closed liv-1382 SW to AFP 1 Suct. Viv Closed IIV-1383 SW to AFP 2 Suct. Viv If the position is changed from normal on the following valves, an unanalyzed condition can be created on a safety system:

Norm. Pos._

Valve No.

Description Closed SG 2 Recirc to Cond.

Closed llV-603 SG 2 Recirc to Cond.

Closed llV-603A SG 1 Recirc to Cond.

Closed liv-611 SG 1 Recirc to Cond.

Closed llV-611A llV-DH63 LPI to HPI 2 Xtie V1v Closed llV-Dil64 LPI to llPI 1 Xtie Viv

COMPLETED FACILITY CHANGE REQUESTS FCR NO:

79-165 SYSTEM: Steam Generator COMPONENT: Feedwater Valves ClfANCE, TEST OR EXPERIMENT:

Fac*ility Change Requet 79-165 was implemented to revise Piping &

Instrument drawing M-007 to show valves FW2685 B&D and FN2686 B&D as normally closed and to change the "Q" boundary so as not to include valve FW2686 E.

This FCR also included the revision of FSK-M-EBD-14-4 and 5 to show the removal of PDS2686A-D and PDS2685A-D respectively.

REASON FOR CHANGE:

These changes were made to update the previously mentioned drawings to reflect "as built" conditions.

SAFETY EVALUATION:

This FCR added to P&ID M-007, the notation that shows valves FW2685 B&D and FW2686 B&D normally closed instead of normally open-In addition, this FCR updated FSK drawings to show the removal of pressure instrumen-tation from these valves. The valves were originally the source valves for pressure switches PDS2685A through D and PDS2686A through D, which indicate steam generator pressure (secondary side).

Due to spurious trips to the SFRCS and plant shutdown, the pressure instruments were moved to a position upstream of the Main Stop Valves (per FCR 77-424).

For this reason the valves FW2685 B&D and FW2686 B&D are not used and are closed to ensure system pressure.

When closed the valves will not affect the safety function of the system. This is not an unreviewed safety question.

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