ML19344F443

Monthly Operating Rept for Aug 1980

ML19344F443
Person / Time
Site:	Davis Besse
Issue date:	09/09/1980
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML19344F442	List: ML19344F441 ML19344F443
References
NUDOCS 8009150277
Download: ML19344F443 (16)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346

- 3.IT Davis-Besse Unit 1 DATE September 9, 1980 Bilal Sarsour COMPLETED BY (419) 259-5000, TELEPHONE -

Extens crf-251 August, 1980-MONTH

. DAY AVER AGE DAILY POh2R LEVEL '

DAY AVERAGE DAILY POWER LEVEL (MWe. Net) (MWe.Ne:1 0 g7 0-1 0 -

gg 0 2

0 0 3 39 0 3 0 4 .

0 0 5 21 0 0 6 22 0 0 7 23 0 0 8- 24 0 25 d

9 0 26 0 10 0 0 11 27 0 0 12 28 0 0 13 29 0 30 0

14 0- 3g 0 15 0

16 .

INS 1 RUCTIONS On this format. list the average daily unit power levelin MWe Net for each day in the reporting month. Compute to the nearest whole megawatt.

. 1 19/77)  !

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE September 9, 198C COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, Extension 251

' OPERATING STATUS Notes DaYis-Besse Unit 1

l. Unit Name:

August. 1980

2. Reporting Period:

2772

3. Licensed Thermal Power (MWt):

925

4. Nameplate Rating (Gross MWe):

906

5. Design Electrical Rating (Net MWe):

934

' 6. Maximum Dependable Capacity (Gross MWe):

890

7. Maximum Dependable Capacity (Net MWe):

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

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

10. Reasons For Restrictions.If Any:

' His Montti Yr.-to-Date Cumulative .

' 744 5,855 26.380

11. Hours In Reporting Period 0 2.078 11.042

12. Number Of Hours Reactor Was Critical 0 0 28,758

13. Reactor Reserve Shutdown Hours 0 2.008.7 11 883 0

14. Hours Generator On-Line '

0 1,728 0

15. Unit Reserve Shutdown Hours 24.886.812 '.

0 4,687,305

16. Gross Thermal Energy Generated (MWH) 0 1,583,559 8.307.070

17. Gross Electrical Energy Generated (MWH) , _ ,

18. Net Electrical Energy Generated (MWH) .

0 1.483.787 7.654.16s 0 34.3 45.8

19. Unit Sen-ice Factor 0 34.3 52.9

20. Unit Availability Factor

- 0 28.5 14.a

21. Unit Capacity Factor (Using MDC Net) 0 28.0 34.2

22. Unit Capacity Factor (Using DER Net) 0 14.3 25.6

23. Unit Forced Outage Rate .

. 24.- Shutdowns Scheduled Over Next 6 Months (Type.Date.and Duration of Eacht: ,

'None~ .

September 15, 1980 r

_ 25. If Shut Down At End Of Report Period, Estimated Date of Startup:

26. Units in Test Status (Prior to Commercial Operation): Forecast Achieved -

INITIAL CRITICALITY INITIAL ELECTRICITY -

COMMERCIAL OPERATION ,

-(Up?)

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"I g, DOCKET NO. 50-346 ' ,

UNIT SHUTDOWNS AND F0WER REDUCTIONS h==. Trait 1 UNIT NAME qavin eptember 9. 1980 g

DATE -f i

COMPLETED IlY Bilal Sarsour_

August. 1980 419-259-sono: Ext. 21 l REPORT MONTil TELEPI!ONE

- i s a 3g Licensee ,g E Eg cause & Corrective . I

-, .$ ? $.Yh

,g g y 5? Action to g 3g Event y!"b

.No. Date 2 F 5. Report n *N0 Prevent Recurrence

$E. j15 g -

0 4 80 04 7 S 744 C 4 NA NA NA The unit outage which began on ,

- April 7, 1980, was still in progress .s '

through the end of August, 1980. See Operational Summarf for further details.

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Exhibit G. Instructions Reason: Method:

F: Forced 1-Manual for Preparation of Data S: Schedu!cd A Equipment Failure (Explain) Entry Sheets for Licensee B Maintenance of Test 2-Manual Scram.

3-Automatic Scram. Event Report (LER) File (NUREG.

C Refueling O+1cG#eid 0161)

D Regulatory Restriction li-Operator Training & License Examination 4-Continuation S

F Administrative 5-Reduction .

Exhibit ! Same Source G Operational Estor (Explain) 6-Other (9/77) II Other (L xplain)

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OPERATIONAL

SUMMARY

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. AUGUST, 1980 ,

e The unit outage,'which began on April .7,1980, was still in progress through the .

end:of August, 1980,~ with hanger modifications required to satisfy NRC Bulletin 79-02 and 79-14 being the critical work. In addition, several snubber mountings required modifications.

The Reactor Coolant System is full and vented, however, the hanger work prevents ,

system heatup.- -

Major l work completed this month was:

1. Completion of the control grade T-sat meter installation

2. DH-11 and DH-12 interlock with the pressurizer heaters

3. Power operated relief valve indication

-4. Installation.of the makeup tank level interlock Training of operators on Facility Change Requests and Technical Specification changes associated with Cycle 2 has been completed.

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DATE: August, 1980 REFUELING INFORMATION Davis-Besse Nuclear Power Station Unit l'

1. Name of facility:

March, 1982

2. Scheduled date for next refueling shutdown:

May, 1982

3. Scheduled date for restart following refueling: _,

4. Will refueling or resump. tion of operation thereaf ter require a technical If answer is yes, what, specification change or other license amendment?

in general, will these be? ~ If. answer is no,.has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to determine whether 'any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)? .

Reload analysis not completed, none identified to date.

5. Scheduled date(s) for submitting proposed licensing action and supporting information. July, 1981 e.g., new or

6. Important licensing considerations associated with refueling, different fuel design or supplier, unreviewed design or perf ormance analysis methods, significant changes in fuel design, new operating procedures.

None identified to date.

7. The number of fuel assemblies (a) in the core and (b) in the spent fuel storage pool. 44,- Spent Fuel Assemblies 8 - New Fuel Assemblies

(,)' 177 (b)

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

735 Increase size by 0 (zero)

Present

9. The projectr a date of the last refueling that can be discharged to the spent

. fuel pool assuming the present licensed capacity.

Date 1988 (assuming ability to unload the entire core into the spent fuel

- pool is maintained) .

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9 COMPLETED FACILITY CHANGE REQUEST FCR NO: .77-244

SYSTD(: Process and Area Radiation Monitors

, COMPONENT: -All Radiation Monitors CHANGE, TEST, OR EXPERIMENT: On April 26, 1980, Victoreen furnished " trip test function" switches which we're installed on all radiation monitors.

REASON FOR THE CHANGE: Previously there was'no method of testing alarm and trip functions and meter response without using test equipment and/or internal (to

-module) switches and the aid of an I&C technician. Because the alarm potentiometers had to be adjusted each time, the installation of the switches reduces the wear of the potentiometers and eliminates the need of an extra man to perform the surveil-lance tests.

SAFETY EVALUATION: The FCR calls for the installation of alarm / trip function test switches. These switches allow testing of alarm and trip functions and meter res-ponse. This will. result in an improvement in the system and will not adversely-affect the safety of_the system.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO.77-516 SYSTEM: Station Instrument Ground COMPONENT: UTR Cabinet (C3921)

CHANGE, TEST, OR EXPERIMENT: FCR 77-516 called for the installation of an instru-rient ground wire from the UTR cabinet (C3921) to the remote analog multiplexer cabi-net (C4601), and-then from the remote analcg multiplexer to the station computer cabine.t D.C. mecca ground. The work was completed on 1.pril 26, 1980.

REASON FOR THE CHANGE: The cables and the UTR plate were not properly grounded'(no ground wire installed) during construction.

SAFETY EVALUATION: This change will provide proper grounding of the UTR plate and cable shields. The entire scope of this FCR is non-nuclear safety related except where the ground wire passes through the containment electrical penetration. This change will provide additional assurance of system operability and will not create a new adverse environment. This change does not constitute an unresolved safety question.-

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COMPLETED FACILITY CHANGE REQUESTS FCR NO.78-010 ,

SYSTEM: Process and Area Radiation Monitoring  !

COMPONENT: Radiation Monitors  ;

4 CHANGE, TEST, OR EXPERIMENT: On February l'9, 1980, needle valves were installed ,

on the high and low pressure sides of the Barksdale flow switches on RE1003A, RE1003B, RE2024, RE2025, RE5029, RE5030, RE5052, RE5327, RE5328, RE5403, and RE5405 REASON FOR THE CHANGE: There has existed a " fluttering" problem on the Barksdale ,

flow switches which caused excessive burning of relays and alarms on the computer.

SAFETY EVALUATION: The needle valves will eliminate the " fluttering", problem by ccting as snubbers on the high and low pressure sides of the flow switches for the i radiation monitors. Two of these valves were placed on RE2024 on November 30, 1977 gnd have proved to be an effective solation to this problem.

This change will improve the operation of the equipment and will not adversely cffect the safety of the plant.

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COMPLETED FACILITY CHANGE REQUEST

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FCR No: 79-143 . .

. SYSTEM: - 4.16.KV Bus C1 (D1) i; COMPONENT: Synchro-Check Relay /25 C1 (D1)

o CHANGE, TEST, OR EXPERIMENT: 0.n December 29, 1979, are suppression diodes were Einstalled and tested on the Westinghouse type CVE synchro-check relays. These i'

wsre installed to enable proper relay operation.

REASON FOR THE CHANCE: During routine preventive maintenance on the synchro-check ralays, the moving contact was found welded to the stationary contact of the relay.

This is a recurring.probles as FCR 79-523 identified the same problem for the 13.8 KV Bus A and.B' synchro-check relays.

- SAFETY EVALUATION: The orig'inally supplied control circuit for the 4160 VAC busses synchronizing scheme' utilized a contact from the CVE relay to drive an auxiliary relay. It was found that welding of the CVE contact occurred, and Westinghouse cdvised that this was due to the low closing torque of the relay which resulted in poor contact wipe.

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One fixL for the stated problen is to place a diode across the auxiliary relay coil.

• Thia diode is 'normally backbiased, but provides a circuit to release the stored

-energy of the coil if the CVE contact were to open. This' circuit eliminates the C

- possibility of the scored energy trying to release across the CVE contact. ,

~ The diode fix-is~ functionally compatible with the existing circuit's operation. It rasolves contact arcing and welding and it therefore enhances system reliability.

-The nuclear' safety of plant personhel and the public is not adversely affected.

- This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS t-i

,FCR NO: .79-166 SYSTEM: Steam and Feedwater Rupture Control System (SFRCS)

COMPONENT: Cabinet C5721 CHANCE, TEST, OR EXPERIMENT: ,0n May 17, 1980, studs were installed on the field wiring side of Terminal Boards 11, 13, 25, and 27, Terminals 13 and 16 in C5721.

REASON FOR THE FCR: During the performance of the SFRCS Monthly Test, these ter-ninals are used to. simulate reactor coolant pumps running when the reactor coolant pumps are off._ Previously alligator type jumpers were placed on the head of the cerews'between terminals.13 and 16. The installation of the studs eliminates the possibility of the jumpers falling off and causing equipment damage and/or personnel injury.

SAFETY EVALUATION: 'This FCR involves the installation of terminal block spacers to prevent equipment damage and/or personnel injury during the testing of the SFRCS.

This change will not adversely affect the function of the SFRCS. It will improve the reliability for performing the SFRCS monthly test. All modifications are inter-nal to the cabinet and will not prevent the safe shutdown of the plant.

This change does not. create any new adverse environments and does not constitute an

- unreviewed safety' question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-213 SYSTEH: Diesel Generators COMPONENT: C-3618 CHANGE, TEST, OR EXPERIMENT: On May 1, 1980, new relays, relay sockets, and diodes wire installed and tested to c'nable proper annunciator action for SCR-Diode Failure Excitor Regulator Alarm.

REASON FOR THE CHANGE: When running the diesel generator, the SCR diode failure clarm is on, however, there are no red lights (which indicate an alarm) on the SCR diodes. Several resistors and relays for this alarm have failed, and it is believed this happens because of a design problem in the circuit.

SAFETY EVALUATION: This ICR consists of installing new relays, relay sockets, and diodes associated with the SCR Diode Failure Exciter Regulator Alarm.

This change will not affect the safety function of the Emergency Diesel Generators.

It will improve the reliability of the annunciator alarm circuit.

The modifications are internal to cabinets C-3617 (C-3618) and will not prevent a cafe shutdown of the plant.

All work involved with this package has been done under the supervision of the vsndor. ' Installation in accordance with vendor's instructions is to insure no cdverse environment is created. Tais is not an unreviewed safety question.

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-COMPLETED FACILITY CHANGE REQUEST

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5 FCR NO: 79-273 .

t SYSTEM: Reactor Protection System (RPS) l' t

COMPONENT: Cabinets C5762,.C5763, C5755, and C5756 -

CHANCE TEST. OR EXPERIMENT: ,0n May 16, 1980, studs with a locknut were installed on the field wiring side of reactor protection system channel location 1-10-6, terminals 3 and 4'in Cabinets C5762, C5763, C5755, and C5756 (RPS Cabinets Channels 1, 2, 3, and 4).

REASON FOR THE FCR: During the performance of ST 5030.12, " Channel Functional Test of the Reactor Trip Module Logic and Control Rod Drive Trip Breakers", an

.c111 gator clip slipped off of one of the terminals causing a short across the assential 120 VAC instrument bus. This in turn tripped inverter YV4 causing de-ener-giration of Safety Features Actuation System Channel 4 and RPS Channel 4 as well as causing the closure of DH-11 isolating the Decay Heat System.

.This FCR will-prevent the recurrence of Licensee Event Report NP-32-79-09 as caused by the slippage of the alligator clip off the flathead screw. This FCR is similar to FCR 77-139 which was implemented on the Steam and Feedwater Rupture Control System.

SAEETY EVALUATION: This FCR provides for the addition of a brass spacer to specific L

RPS terminal strip lugs. This change will provide the ability to securely attach En alligator clip.without s11ppL.g and the consequent possible shorting of other terminal strip lugs. All changes are-internal to the RPS cabinets and will not create a new adverse environment. .Tbis change will provide additional assurance cf system operability.and testability.- This change does not constitute an unresolved cafety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-382 SYSTEM: Main Steam COMPONENT: Restraint W3 CHANGE, TEST, OR EXPERIFENT: On April 19, 1980, work was completed for FCR 79-382 (including Supplement 1). This FCR replaced the bolts on the main steam restraint W-3 and installed them with a verified gap.

REASON FOR THE CHANCE: The existing installation did not permit the bolts to move in elongated ho es upon heatup.

SAFETY EVALUATION: This change involved replacing bolts and installing with a veri-fled gap such that a bolt can move in clongated holes on restraint. This will permit the restraint to function as designed without overstressing the bolts. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO.79-416 SYSTEM: Process and Area Radiation Monitoring System COMPONENT: Station Vent Stack Monitors, RE-2024 and RE-2025 CHANCE, TEST, OR EXPERIMENT: 'On March 11, 1980, work was completed on FCR 79-416.

This FCR changed the power to the motors for station vent stack vacuum pumps 1-1 cnd 1-2 en RE-2024 and RE-2025 to MCCE12A and MCCF12A, respectively.

REASON FOR THE CHANCE: According to the October 30, 1979 letter on lessons learned requirements from Mr. Denton of the NRC, item 2.1.8.b states, "If normal AC power is used for sample collection and analysis equipment, an alternate backup power supply should be provided." Since these pumps are used for sample collection, the motors should be supplied with power from the diesel generator in the event of a loss of normal AC power. The motor for pump 1-1 on RE-2024 had channel A power and was changed to channel 1 power, and the motor for pump 1-2 on RE-2025 had channel B power and was changed to channel 2 power.

SAFETY EVALUATION: The station vent stack monitors RE-2024 and RE-2025 are two iden-tical systems. The system is designed so that an isokinetic sample is always being withdrawn from the station vent stack.

The station vent stack monitors consist of three channels; station vent particulate monitor (RE-2024A, RE-2025A), station vent iodine (RE-2024B, RE-2025B) and station vent gas monitor (RE-2024C, RE-2025C). By upgrading the power supply to Class IE, .

the station vent stack monitors normal functions are not affected. This change will facilitate the operation of the monitors during a loss of offsite power, since the monitors will be supplied from essential buses E12A and F12A. This change will not affect the Class 1E source because the power circuit for these monitors can be separ-ated from the 480 volt motor control centers E12A and F12A by Class 1E c*rcuit breakers.

Ground fault protection 10 existing on the neutral of the transformer in Unit Substa-tions El and F1 which supplies power to the 480 volt motor control centers E12A and F12A. Therefore, an unreviewed safety question does not exist.

O COMPLETED FACILITY CHANGE REQUESTS FCR NO.80-028 SYSTEM: Clean and Miscellaneous Liquid Radwaste COMPONENT: Various CHANGE, TEST, OR EXPERIMENT: FCR 80-028' changes the system design criteria from ASME Section III Class C or Class III to the design criteria in the following docu-rents.

1) Branch technical position ETSB 11-1, Rev. 1

2) 10 CFR20 Parts 20.1 (c) and 20.106

3) 10 CFR50 Appendix A, general design criterion 60 REASON FOR THE CHANGE: The systems as designed now exceed the current NRC require-cent for radwaste systems. Declassifying the systems from ASME Section III to meet these requirements (shown above) will facilitate the making of necessary modifications cud design changes as well as being of economic benefit to Toledo Edison.

SAFETY EVALUATION: The present design criteria for major processing equipment in the liquid radioactive waste systems are ac follows:

A. The individual isotopic concentrations in liquid effluents at the site boundary shall not exceed the limits fer releases to unrestricted areas given in Appendix B of 10CFR Part 20.

B. The releases. of radioactivity f rom the station shall comply with the "as low as practicable standard set forth in 10CFR Part 50.34(a) & 10CFR Part 20.1(c).

Major processing components in the liquid radioactive waste systems (Lar' .g the piping) conform to the requirements of ASME Section III Class III or Clas In the Safety Evaluation Report (NUREG 0136) for Davis-Besse Unit 1, the NRC 2 aviewed the liquid radwaste systems on the basis of Franch Technical Position ETSB 11-1, Rev. I " Design Guidance For Radioactive Waste Management Systems Installed In Light-Water Cooled Nuclear Power Plants",10CFR20 1(c), General Design Criterion 60 and 10CFR20.106. Since the releases of radioactive materials from the system met the requirements of 10CFR20 and the system design met the requirements of BTP ETSB 11-1, Rev. 1, the NRC found the system to be acceptable.

Future modifications to the liquid radwaste systems shall be made in accordance with the criteria used to perform the Safety Evaluation Report. That is, future modifica-tions shall meet the requirements of the following:

1. BTP ETSB 11-1, Rev. I dated 11/14/75

2. 10CFR20

3. Cencral Design Criterion 60 (10CFR50 Appendix A)

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e COMPLETED FACILITY CHANGE REQUESTS FCR 80-028

• PAGE 2 This modification of the original design criteria (ASME Section III), which meets ..

the design requirements deemed acceptable by the Safety Evaluation Report, does not increase the probability of occurrence or the consequences of an accident or mal- ,

function of equipment important to safety as previously evaluated in the Safety Analysis Report. In addition, this modification does not create the possibility

• of an accident or malfunction of a different type than previously evaluated, or reduce the margin of safety as defined in the basis for any Technical Specification.

The require-Therefore, this change does not involve an unreviewed safety question.

ments listed above (1,2, & 3) formed the basis for the NRC's evaluation of the liquid radwaste systems at Davis-Ecsse, so, in ef fect, this is a specifically reviewed safety question.

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