ML19210F002

Monthly Operating Rept for Nov 1979

ML19210F002
Person / Time
Site:	Davis Besse
Issue date:	12/07/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19210E999	List: ML19210E998 ML19210F002
References
NUDOCS 7912130375
Download: ML19210F002 (13)

Text

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50 '146 DOCKET NO. Davis-Besse Un_it 1 UNIT S!!UTDOWNS AND f OW::.:t REDUCTIONS UNIT N AME November 9.

1979 DATE Frdal Caba COM;'LET ED BY REPORT MONTl! October. 1979 TELEPil0NE 419-259-5000 Ext. 236 l

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E Cause & Corrective

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No.

Date g

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15 79 10 5 S

48.8 A

1 NA CJ VALVEX Shutdown to repair pressurizer spray valve RC 2.

16 79 10 15 F

142.2 A

3 NP-32-79-ll HA INSTRU Capacitor failure in Integrated Con-trol System (ICS) pulser circuit to the turbine electro-hydraulic control system. Refer to attached summary for further details.

i 17 79 10 25 F

156.4 A

3 NP-33-79-121 CB CKTBRK Loss of Reactor Coolant Pump 2-2 from b

blown fuse in the DC power supply b

9 starting a pump two minute time delay trip relay with Reactor Coolant Pump W

1-1 already shutdown.

6 9

-a M

C k cd -

w to w

N cc b

O Exhibit G Instructions Method:

for Preparation of Data

[3 '

F: Forced Reason:

1 Manual S: Schedu!cd A. Equipment Failure (Explain) 2-Manual Scram.

Entry Sheets for Licensee Event Report (LER) Fife (NUREG-B. Maintenance of Test 3 Automatic Scram.

Ln C. Refueling 40the: (Explain) 0161)

Ln D. Regulatory Restriction gp 1:. Operator Training & License Examination S

U F. Ad nu nist rat ive Exhibit ! - Same Source G Opei.itional I nor (Explain) u (9/77) h II Other (l silain)

N

AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

Davis-Besse Unit 1 W'IT December 7, 1979 DATE Erdal Caba COMPLETED BY 419-259-5000, Ext.

TELEPiiONE 236 November, 1979 MONTil DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Neti (MWe-NetI 0

O g7 I

0 0

jg 2

0 0

39 3

0 4

0 20 555 0

21 5

660 0

22 6

636 0

23 7

649 0

24 8

609 0

25 9

595 0

26 10 605 0

27 11 443 0

28 12 332 0

29 13 209 0

30 I4 0

3g IS 0

16 INSTRUCTIONS On this format.hst the aserage daily unit power level in MWe Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77) 1551 348

OPERATING DATA REPORT 50-346 DOCKET NO.

December 7, 1979 DATE COMPLETED BY Erd Caba TELEPilONE 419-

- 000, Ext.

236 OPERATING STATUS N"I'5 Davis-Besse Unit 1

1. Unit Name:

2. Reporting Period:

November. 1979 2772

3. Licensed Thermal Power (MWt):

925

4. Nameplate Rating (Gross MWe):

906

5. Design Electrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe): to be determined to be determined

7. Maximum Dependable Capacity (Net MWe):

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

None

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

10. Reasons For Restrictions,if Any:

This Month Yr..to.Date Cumulative 720 8,016 19,781

11. Ilours In Reporting Period 259.8 4,303.3 10,935.1

12. Number Of flours Reactor Was Critical 0

2,085.5 2,875.8

13. Reactor Reserve Shutdown flours 241.0 4,141.6 9,874.8 14, llours Generator On-Line 0

1,728.2 1,728.2

15. Unit Resene Shutdown flours 409,367 10,011,131 20,198,701

16. Gross Thermal Energy Generated (MWii) 137,67_8 3,339,756 6,723,511

17. Gross Electrical Energy Generated (MWII) 121,083 3,129,118 6,170,578

18. Net Electrical Energy Generated (MWil) 33.5 51.7 51.5

19. Unit Service Factor 33.5 73.2 61.3

20. Unit Asailability Factor to be determined

21. Unit Ccpacity Factor (Using MDC Net) 18.6 43.1 37.7

22. Unit Capacity Factor (Using DER Net) 66.3 18.1 23.2

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Next 6 Months (Type.Date.and Duration of Eacht:

Refueling Outage, March 1980 12 weeks December 11, 1979

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

Forecast Achiesed

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

INITIA L CRITICALITY INITIAL ELECTRICITY COMMERCIA L OPFR ATION 1551 349

~

November, 1979 DATE:

RERIELING INFORMATI01 Davis-Besse Nuclear Power Station Unit 1 1.

Name of facility:

March, 1980 2.

Scheduled date for next refueling shutdown:

June. 1980 Scheduled date for restart following refueling:

3.

Will refueling or resumption of operation thereaf ter require a technical If answer is yes, what, 4.

specification change or other license amendment?If answer is no, has the reload fue in general, will these be?

Safety Review Cc=mittee and core configuratica been reviewed by your Plant to determine whether any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached

~

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

information.

D ec ember, 1979 Important licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or performance analysis 6.

nethods, significant changes in fuel design, new operating procedures.

fuel pool capacity expansion program was approved by the NRC The spent in Amendment 19 to the operating license received August 1,1979.

The number of fuel assemblies (a) in the core and (b) in the spent fuel 7.

storage pool.

0 (zero) 177 (b).

(a)

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, 8.

in number of fuel assemblies.

Present 735 Increase size by 0 (zero)

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

fuci pool assuming the present licensed capacity.

Date___1989 (assuming ability to unload the entire core into the spent fue pool is maintained and the unit goes to an lo month retueling cycle) 155i 350

t 50-346 DO Davis-Besse Unit 1 UNIT S!!UTDOWNS AND POWE:t REDUCTIONS uI E

December 7, 1979

)

DATE Erdal Caba COMPLETED BY REPORT MONTil November, 1979 TELLPl!CNE 419 ')So-5000 - Ext. 236 "E

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• vt Cause & Corrective E

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Licensec

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Ai. tion to No.

Date

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Prevent Recurrence f:

35F Report a H

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Loss of Reactor Coolant Pump (RCP) 17 79-10-25 F

474.2 A

3 NP-33-79-121 CB CKTBRK 2-2 from blown fuse in DC power sup-(Continued)

ply, Replaced all four RCP seals.

See Operational Summary for details of further work items completed dur-<

ing the outage.

18 79-11-28 F

0.0 A

4 NA NA NA Reactor power was reduced to approxi-i mately 50% to run another heat balanc(

to get the NIs in tolerance when the

('

S power range NIs were reading slightly greater than 2% below the calculated M

heat balance.

(Q Absolute Position Indication (API) fo i

r_

19 79-11-29 F

0.0 A

4 NA NA NA Group 7, Rod 5 with API Group 5, Rod 11 previously declared inoperable was

[gc) declared inoperable initiating a re-ddction in reactor power to a 42%.

W N

L.n b

4 3

Exhibit G -Instructions i

2 Method:

F: Forced Reason:

1-Manual for Preparation of Data U

S: Schedu!ed A-Equipment Failure (Explain) 2 Manual Scram.

Entry Sheets for Licensee W

B. Maintenance of Test 3 Automatic Scram.

Event Report (LER) File (NUREG-C-Refueling 4-Other (Explain) 0161)

~

D Regulatory Restriction I'.-Operator Training & License Examination 5

F-Administ ratis e Exhibit 1 - Same Source G-Operational Eiror (Explain)

(9/77) iI Other (1:xplain)

1 I

50-346 DOCKET NO.

UNIT S!!UTDOWNS AND IOWER REDUCTIONS UNIT NAME Davis-Bes'io Unit 1 i

DATE Doro-hor 7 1479 f

COMPLETED BY Erdal Caba I

REPORT MONT!!' November, 1979 TELEPil0NE 419-259-5000. Ext. 236 l

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1 E

3

)Y Cause & Corrective i

Licensee

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$*3 o

9-A'"0" '"

No.

Date E.

5E s

.5 s 5 Event 17 C

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C m= =s Report u mu 8O Prevent Recurrence c-2 v

6 Maintenance outage due to a low be'ar-20 79-11-30 S

4.8 B

1 NA NA NA ing oil level alarm on RCP 1-2.

See Operational Summary for further de-tails.

eY i

6 9

6 9

m 6

0 E559 pa V4 PD b

'M 3

4 i

Method:

Exhibit G Instructions W

F: Forced Reason:

1-Manual for Preparation of Data S: Schedu!ed A Equipment Failure (Explain)

~

B Maintenance of Test 2-Manual Scram.

Entry Sheets for Licensee 3 Automatic Scram.

Event Report (LER) Fife (NUREG-C Refueling LM D Regulatory Restriction 4-Other (Explain) 0161)

N l' Operator Training & License Examination 5

F Adminiuratise Exhibit I - Same Source G-Operational lit ror ( fixplain)

(9/77)

!! Other (laplain)

OPERATIONAL

SUMMARY

NOVEMBER, 1979 The unit shutdown which was initiated on October 25, 1979, when Reactor Coolant Pump 20, 1979, (RCP) 2-2 tripped f rom a blown f use, continued until 1811 hour0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br /> s on November Below is a list of the major work items when the turbine-generator was synchronized.

performed during the outage:

1.

All four RCP seals were replaced.

A new expansion joint for 1-4-2 heater extraction was rewelded.

2.

A design change to the Couch relays in the RCP starting interlock circuits 3.

This initiated a similar change in sixteen safety related cir-was made.

cuits.

4.

NI-3 was pulled and a new detector was installed. New pre-amp temperature cable was installed for NI-1.

5.

Several. pipe restraints were modified per IE Bulletin 79-14.

11/20/79 - 11/21/79 The unit returned on line at 1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br /> or. November 20, 1979, and reactor power was increased to 81% of full power with generator gross load at 760 MWe on November 21, 1979.

11/22/79 Reactor Coolant System (RCS) flow inidcation was determined to be less than required by Technical Specification 3.2.5 at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />.

Reactor power was decreased to 73% at which point RCS flow indication was back within allowable limits. At 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> generator gross load was 684 MWe.

11/23/79 - 11/25/79 Reactor power was maintained at approximately 73% until 0620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br /> on November 25, 1979, when RCS flow inlication was again determined to be less than that required by Technical Specifi-cations. Reactor power was decreased to 67% at which point the RCS flow indication was again back withir. tolerance. The cause of the low flow indication is still being investigated.

9 11/26/79 Reactor power was maintained at approximately 69% of full At 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br /> on November 26, 1979, RCP 1-2 was manu-power.

ally tripped due to a lower motor bearing low oil level alarm.

11/27/79 - 11/28/79 The unit remained at approximately 68% of full power with generator gross load of 635 MWe until 0645 hours0.00747 days <br />0.179 hours <br />0.00107 weeks <br />2.454225e-4 months <br /> on November 28, 1979, when it was determined that the power range nuclear instrumentation (NIs) were reading slightly greater than 2%

below the calculated heat balance.

Reactor power was reduced to 50% to bring the NIs in tolerance and to run another heat balance to get the NIs in tolerance.

1551 353

OPERATIONAL

SUMMARY

NOVEMBER, 1979 PAGE 2 0F 2 11/29/79 The Absolute Position Indication (API) for Group 7, Rod 5 was declared inoperabic at 0640 hours0.00741 days <br />0.178 hours <br />0.00106 weeks <br />2.4352e-4 months <br /> with Group 5, Rod 11 API pre-viously declared inoperable, initiating a reduction in reactor power to approximately 42% and resetting the high flux trip sepoint to 54%.

11/30/79 Reactor power was reduced and the turbine-generator taken off line at 1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br /> to investigate the lower motor bearing oil 1cyc1 alarm on RCP 1-2; to fix Group 7, Rod 5, and Group 5, Rod 11 API.

1551 354 en

REFUELING INFORMATION Continued Page 2 of 2 4.

The following Technical Specifications (Part A) will require revision:

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System Instrumentation Setpoints (and Bases) 3.1.3.6 - Regulating Rod Insertion Limits 3.1.3.7 - Rod Program 3.2.1 - Axial Power Imbalance (and Bases)

The following Technical Sepcifications (Part A) may also require revision:

3.1.2.8 & 3.1.2.9 - Borated Water Sources (and Bases) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - DNB Parameters (and Bases) 1SL51 355

FACILITY CHANGE REQUEST COMPLETED DURING NOVEMBER, 1979 s

l l

FCR NO: 77-094 Post Accident Containment Airborne Radiation Monitors RE 5029 and RE 5030 SYSTEM:

COMPONENT:

Pump interlocks CHANGE, TEST, OR EXPERIMENT _: On June 11, 1979, installation and testing of a low flow trip interlock on both post accident containment airborne radiation monitors RE 5029 This interlock trips the sample and RE 5030 was completed as requested by FCR 77-094.

This change was made pump of the radiation monitor should a low flow condition exist.

with the recommendation of the unit architect / engineer, Bechtel Company.

and outlet valves for RE 5029 and RE 5030 as REASON FOR THE FCR: The sample inlet isolation valves.

well as the containment hydrogen analyzers are in fact containmentthese valves would close Upon receipt of a Safety Features Actuation System actuation, The interlock depriving the sample pumps of suction as well as deadheading them.thus preventing pump which was added trips the sample pumps of the radiation mot.itors, damage.

those pumps are in Review of the sample pumps of the hydrogen analyzers revealed that fact vacuum pumps which are capable of being operated with the valves closed without Therefore, no modification was decmed necessary on this system.

damage.

The proposed modification would provide radiation monitors RE 5029 SAFETY EVALUATIO!{:

to stop the sample pumps in such cases as closure and RE 5030 with low flow interlocks The addition of the circuitry necessary to accomplish this does of the inlet valves.

the proper functioning of the radiation monitoring system during its re-not affect A

quired use as post accident instrumentation.

The modification serves to enhance the probability that the equipment will be availa-ble to perform its intended function.

1551 356 t

FACILITY CHANGE REOUEST C0MPLETED DURING NOVEMBER, 1979 FCR NO: 79-246 SYSTEM: Auxiliary Feedwater System COMPONENT: Mode selector switches HIS520B and HIS521B CHANGE, TEST, OR EXPERIMENT: On June 23, 1979, installation of mechanical interlocks (stops) on the mode selector switches HIS520B and HIS5213 as requested by FCR 79-246 was completed. These switches, located on control room center console, select the control mode of the Auxiliary Feedwater Pump Turbines. The mechanical stops which were added prevent movement of the switches into the Integrated Control System con-trol mode.

REASON FOR THE FCR:

The stops were added to comply with the Nuclear Regulatory Commission order dated May 16, 1979, Item IV(1)(b).

SAFETY EVALUATION: The addition of this mechanical interlock will only prevent Integrated Control System control of the Auxiliary Feedwater System. The manual and auto-essential (both safety grade) control functions will not be changed in any way.

Therefore, the safety function of the Auxiliary Feedwater System will not be affected. This is not an unreviewed safety question.

1551 357

FACILITY CHANGE REQUEST COMPLETED DURING NOVE!EER, 1979

-. ~'

FCR No: 78-074 SYSTEM: Service Water COMPONENT: Flow Elements (FE) 9808 and 9809 CHANGE, TEST, OR EXPERIMENT: On October 4,1979 the physical work and inspections associated with the installation of 1/2" pipe nipples with threaded pipe caps on the source valves of FE 9808 and FE 9809 were completed.

FE 9808 and FE 9809 are the service water flow elements in the outlets of control room emergency condensing units 1-1 and 1-2, respectively.

REASON FOR THE FCR:

The threaded nipples were added in order to provide a means of attaching a local dif ferential pressure gauge to the flow elements for calibration purposes.

SAFETY EVALUATION:

The change involves only the addition of threaded nipples to the source valves of the Q-listed flow elements FE 9808 and FE 9809. The addition of the nipples will not affect the operation of the flow element.

An unreviewed safety question does not exist.

155l 358

FACILITY CHANGE REQUEST COMPLETED DURING NOVEMBER, 1979 en FCR NO: 78-503 SYSTEM: Radiation Monitoring COMPONENT: Radiation Monitor RE 8432 CHANGE, TEST, OR EXPERIMENT: On October 20, 1979 modifications to the discharge pip-RE 8432 ing of radiation monitor RE 8432 were completed as requested by FCR 78-503.

to an alarm only; it monitors the service water discharge header to provide an input does not control any equipment. The specific change was to re-route the monitor dis-charge to a floor drain via a globe valve rather than back into the service water piping.

REASON FOR THE FCR: With the former arrangement insufficient, differential pressure occurred across the radiation monitor during some modes of service water system opera-This resulted in the inability to maintain the correct sample flowrate at all tion.

times.

The work involves rework of 3/4" - HBC - 36, an existing equipment SAFETY EVALUATION:

The addition of the 3/4" globe valve (SU 8432G) in the non-Q, non-seismic drain.

portion of this line to the equipm,ent drain will not adversely affect the operation This is not of RE 8432. The line is non-essential and not required for safe shutdown.

an unreviewed safety question. RE 8432 is included in the Environmental Technical Specifications, License Appendix B.

I551 359

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