ML19326A249

Monthly Operating Rept for Aug 1978

ML19326A249
Person / Time
Site:	Davis Besse
Issue date:	09/08/1978
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19326A247	List: ML19326A246 ML19326A249
References
NUDOCS 8002030109
Download: ML19326A249 (11)

Text

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AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

g.g7 Davis-Besse Unit 1 September 8 _1978 DATE COMPLETED BY Erdal C. Caba TELEP110NE 419-759-5000. Ext.

236 t

MONni Aucust. 1978 i

DAY AVERAGE DAILY POWER LEVE7 DAY AVERAGE DAILY POWER LEVEL (MWe-Ne:)

(MWe-Net) 0-266 37 gg 0

107 '

2 0  ;

3 300 39 O l 4 . 300 20 l

307 2i 5

308 22 6

0 304 3 7

0 303 ,3 8

306 25 9

0 10 310 26 0

gg 356 7 414 28 12 627 29 13 428 351 14 30 663 0 33 15 0

16 INSTRUCTIONS On this format. list the average daily unit power levelin MWe Net for each day in the reporting month. Compute to the ricatest whole megawaii. .

(9/77 n 8002030{oj

O OPERATING DATA REPORT

. DOCKET NO. 50 't 4 6 DATE September 8, 1978 COMPLETED BY Erdal C. Caba TELEPHONE ^ 19,190 9000, Ex t.

236 OPERATING STATUS Davis-Besse Ur it 1 Notes

1. Unit Name:

2. Reporting Period: Angnee 147A

3. Licensed Thermal Power (MWt): 2772

4. Nameplate Rating (Gross MWe): 925

5. Design Electrical Rating (Net MWe): 906

6. Maximum Dependable Capacity (Gross MWe):

to be determined t be dete M nd

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): 906

10. Reasons For Restrictions. If Any:

This Month Yr. to-Date Cumulative II. Hours In Reporting Period 744 5831 8836

12. Number Of Hours Reactor Was Critical 383.5 2917.2 4709.3

13. Reactor Reserve Shutdown Hours 0 19.8 403.5

14. Hours Generator On-Line 360.2 2615.1 4082

15. Unit Reserve Shutdown Hours _0 0 0

16. Gross Thermal Energy Generated (MWH) 469.519 4,639,135 6,303,167 .

17. Gross Electrical Energy Generated (MWH) 151,451 , _ ,

1,580,173 2,104,622

18. Net Electrical Energy Generated (MWH) 129,032 1,433,946 1,863,764

19. Unit Service F.setor 48.4% 44.8% 46.Z;

20. Unit Availability Factor 48.4% 44.8% 46.2%

21. Unit Capacity Factor (Using MDC Net) to be determined

22. Unit Capacity Factor (Using DER Net) 10 1 Y 27.1% 23.3%

23. Unit Forced Outage Pate 51.6% 31.8% 27.4%

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

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

26. Units in Test Status (Prior to Commercial Operation): Forecast Achiesed I INITIAL CRITICALITY 8/12/77 l INITIAL ELECTRICITY 8/28/77 11/ * #7*

I COMM ERCIA L OPERATION

! 12/19/77**

• Declared operational at 25% 1/23/78***

l ** Declared operational at 40% (from 25%) 7/31/78****

• • • Declared operational at 75% (from 40%) N77I l **** Declared operational at 100% (from 75%)

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DOCKET NO.

50-346 ~

UNIT SilUIDOWNS AND F0WER REDUCTIONS* Davis-Besse Unit 1 UNIT NAME _ September 8, 1978 DATE

• Chnrios M- Aim }

August. 1978 COMPLETED BY '7.19-259-5000, Ext. 251 REPORT MONTil TELEPilONE l

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"E c

._ *g 3 y$ Licensec Eg g*L Cause & Corrective No. Date E 3g s 3g& Event M mV Q ^'tio" '"

Prevent Recurrence

1. ?i o:5 g t; g c Report a o U

2 f 6

20 Cont'd F 0.9 11 4 N/A N/A N/A See Report for July of 1978 frora July .

12.9 3 N/A. N/A N/A The unit tripped on' reactor coolant 21 78 08 02 F 11 system low pressure when the feed-water flow was increased too much by

' the Integrated Control System (ICS).

The feedwater flow over-compensation occurred when the control rods s 're

, placed in automatic operation af.er the performance of control rod e ercis a testing. To prevent recurrence of thi s event, the ICS has been tuned for better response and operation practice has been changed, to place feedwater control in manual when the control rods are placed in manual, until 3 4 I 2 Method: Exhibit G Instructions F: Forced Reason: fin Pieparation of Data i A liquipment Failuie(Explain) 1-Manual S: Schedu!cd. 2 Manual Scram. Entry Sheets for Licensee 11 Maintenance oi Test

• 3 Automatic Scram. Event Repost (LER) File (NUREG-C.Rcfueling 0161) 4-Other (Explain)

D Regulatusy Restaletlun E Operatus Training & License Examination 5

' * * * ' "

• F:Adnimist ative - Exhibit t Same Source G-Operational Eirur (Explain) ll Other (lixplain)

(9/77) 1 s .

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DOCKET NO. 50-346 UNIT SilUIDOWNS AND POW 1.lt REDUCTIONS ifNIT NAMEhptemberDavis-Besse Unit 1 8, 1978 DATE COMPLETED BY Charles N. Alm REPORT MONTil A"S" * , 1978 TELEPilONE 419-259-5000. Ext. 251 "t. '

- ,$g 3 Y Licensee ge,,  !% Cause & Corrective .

Action in No. Date g Eg $ 2g2 Event 3,1 h"g

$ Report a mo g Prevent Recurrence t-

[% g p,

• 6 the temperature coefficient becomes

• negative.

A 1 NP-32-78-10 BA INSTRU The unit was shutdown when two of the 22 78 08 14 F 370 four Safety Features Actuation System [

channels were declared inoperable.

i The outage was lengthened when reac .

tor coolant pump 1-1 seal operation ,

malfunctioned. Refer to the attached Operational Sununary for further details.

3 4 I 2 Method: Exhibit G. Instructions F: Forced Reason: for Preparation of Data 1-Manual S: Schedu!ed A. Equipment Failure (Explain) Entry Sheets for Licemce 2 Manual Scram.

B Maintenance of Test 3-Automatic Scram.

Event ReporI(LER) File (NUREG.

C Refucting 4 Otler (Explain) 0161)

' D Reguhtory Restriction E-Operator Training & License Examination 5 *

. F. Administrative Exhibit I .Same Source G. Operational Error (Explain) pf77) il-Other (Explain) 4 g

_ _ _ _ _ _ _ ___ _ _ _ _ _ .m _ __ _

OPERATIONAL

SUMMARY

FOR AUGUST, 1978 8/1/78 The turbine-g> >::ator was synchronize'd on line at 0052 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> after the ventilation dampers to the reactor cavity had been verified open. At 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />, reactor power attained 40 per-cent with e-aerator gross output at 350 + 10 MWe.

8/2/78 The reactor tripped at 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br /> because of reactor coolant

!- system low pressure. The low pressure was caused by feed-water over-compensation when control rod exercise testing was conducted.

Reactor criticality was re-established at 1911 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.271355e-4 months <br />, and the turbine-generator was synchronized on line at 2243 hours0.026 days <br />0.623 hours <br />0.00371 weeks <br />8.534615e-4 months <br />.

8/3/78 - 8/11/78 From the time of synchronization, reactor power was increased and attained 40 percent at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br /> on August 3, 1978. The reactor power level was then maintained at 40 + 2 percent un-til 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br /> on August 11, 1978. During this time period, the generator gross output was at 350 + 10 MWe except when the following conditions occurred.

l At 2230 hours0.0258 days <br />0.619 hours <br />0.00369 weeks <br />8.48515e-4 months <br /> on August 6, 1978, the test TP 800.23, " Unit

' Load Transient Test" was initiated which caused generator l

gross output to vary between 230 and 380 MWe until 0315 hours0.00365 days <br />0.0875 hours <br />5.208333e-4 weeks <br />1.198575e-4 months <br />

.when reactor power was returned to 40 percent. Then again at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> on August 6, 1978, the transient testing was re-sumed and the generator gross output varied between 320 and

.350 MWe until 1320 hours0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br /> on August 6,19.7 8 when the test was completed .

The final generator load variation was caused at 2340 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.9037e-4 months <br /> on August 8,1978 when reactor power decreased because of a loose circuit board in the Main Feedwater Pump 1-1 speed controller. The Main Feedwater Pump 1-2 was started and

• reactor power was returned to 40 percent at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on August 9, 1978. The generator gross output had decreased to 240 MWe before the Main Feedwater Pump 1-2 had stabilized plant conditions.

The 40 percent power level was maintained to complete physics testing which was required prior to operation above 40 per-cent. This testing was prolonged by problems with the opera-tion of the reactimeter, alarm typer and computer.

l 8/11/78 At 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br /> power escalation to 65 percent was initiated with integrated control system (ICS) tuning performed in con-junction. ICS tuning at the 60 percent power level was com-pleted at 2015 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.667075e-4 months <br />. Reactor power was then reduced to 50 l

• ' = " " - ,--r - . . _ -, , , . _ , , , , _ , , _ __ ,_

OPERATIONAL

SUMMARY

FOR AUGUST, 1978 PAGE 2 0F 2 percent because of intermittent " Exhaust Casing High Level" alarms on the Main Feedwater Pump 1-2. The generator gross output at 60 percent was 570 i 10 MWe.

8/12/78 Reactor power was maintained at 50 percent except when power was reduced to 40 percent between 1820 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.9251e-4 months <br /> to 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> to perform additional .'".S euning. The generator gross output at 50 percent was 475 i 10 MWe.

8/13/78 Power escalation to 75 percent was resumed at 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and achieved at 0534 hours0.00618 days <br />0.148 hours <br />8.829365e-4 weeks <br />2.03187e-4 months <br />. The generator gross output was 700

+ 10 MWo.

8/14/78 Reactor power was maintained at 75 percent until 1254 hours0.0145 days <br />0.348 hours <br />0.00207 weeks <br />4.77147e-4 months <br /> when a unit shutdown was initiated because the Safety Features Actuation System radiation channels 1 and 4 were both inopera-ble. Channel 4 had been inoperable since August 7, 1978 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.

The turbine-generator was taken off line at 1342 hours0.0155 days <br />0.373 hours <br />0.00222 weeks <br />5.10631e-4 months <br />.

8/14/78 - 8/29/78 The Safety Features Actuation System (SEAS) Channels 1 and 4 were inoperable because of the failure of radiation elements 2004 and 2007 respectively. The high ambient temperature at these detectors was attributed to be the cause of their f ailure.

The detector RE 2004 was declared operable at 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br /> on August 15, 1978. The detector RE 2007 was declared operable at 0510 hours0.0059 days <br />0.142 hours <br />8.43254e-4 weeks <br />1.94055e-4 months <br /> on August 16, 1978.

The unit shutdown was continued because of L= proper seal .

operation discovered on the Raactor Coolant Pump 1-1 at 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> on August 16, 1978. Due to the outage continur--

tion for seal repair, duct work was also constructed to provide the detectors RE 2004 and RE 2007 with a cooling sys-tem. Other repair work was also initiated but the above items were the major purpose of the outage.

8/29/78 Reactor criticality was declared at 0729 hours0.00844 days <br />0.203 hours <br />0.00121 weeks <br />2.773845e-4 months <br />, and the turbine-generator was synchronized on line at 2340 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.9037e-4 months <br />.

8/3./78 Reactor power had been increased to 75 percent at 1920 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.3056e-4 months <br /> with the generator gross output at 700110 MWe.

8/31/78 Reactor power was maintained at 75 percent power for the performance of required physics testing at this power level.

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FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1978 FCR NO: 77-297 .

SYSTEM:- Auxiliary Shutdown Panel COMPONENT: F.ase Blocks for LT-SP9B3, LT-SP9A4, LT-RC14-3, LT-SP9A3, LT-SP9B4, LT-RCl4-1 CHANGE, TEST, OR EXPERIMENT: FCR 77-297, which was completed on 5/3/78, rerouted the wiring on the fuse blocks for the following level transmitters:

LT-SP9A3 Steam Generator 2 Startup Range Level LT-SP9B3 Steam Generator 1 Startup Range Level LT-SP9A4- Steam Generator'2 Startup Range Level LT-SP9B4 Steam Generator 1 Startup Range Level LT-RC 14-1 Pressurizer Level LT-RC 14-3 Pressurizer Level The wiring was rerouted to the opposite side of the fuse blocks in such a manner as to allow the fuses to be deenergized when the fuse blocks are pulled.

i REASON FOR THE FCR: The fuse holder .ns formerly wired such that, after opening the fused disconnect, the fused end was still energized. This was a personnel hazard when pulling the fuses.

SAFETY EVALUATION: There is no difference in the electrical function between existing cud proposed conditions for the case of fused disconnect closed. For this case, nuclear safety aspects are identical.

However, for the case of fused disconnect open, there is a difference in the electrical function between existing and proposed conditions. Nuclear safety is identical, but personnel safety is improved by having the fused end de-energized.

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FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1978 FCR NO: 77-428 SYSTEM: High Pressure Injection (HPI)

COMPONENT: HPI Line 1-2 vent valve, HP-72 CHANGE, TEST, OR EXPERIMENT: On June 29, 1978, physical work and testing were completed on FCR 77-428. This FCR improved the accessibility of vent valve HP-72. t The second valve of the vent was relocated near the floor of Mechanical Penetra-tion Room #1. The first valve of the vent is to be left open. All the affected drawings were revised by Bechtel. The piping was pneumatically tested with a nitrogen pressure of 3000 psig.

REASON FOR THE FCR: The valves were located on top of HPI Line 1-2, 15 feet above the floor of Mechanical Penetration Room #1. The position of the valves (two valves in series) made access nearly impossible and dangerous. The valve must be operated monthly to vent the line for Surveillance Test ST 5051.01, "ECCS Monthly Test", in order to meet Technical Specification 4.5.2b.

SAFETY EVALUATION: The lowering of the vent valve on the discharge of HPI Pump #1 will not degrade the nuclear safety function of the valve or the HPI System. The lowering of this valve will increase the operator's safety while operating the valve.

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m FACILITY CHANGE REOUESTS COMPLETED DURING AUGUST, 1978 FCR NO: 77-462 and 78-031 .

SYSTEM: Safety Features Actuation System (SEAS)

COMPONENT: Containment Radiation Monitors RE-2004, RE-2005, RE-2006, RE-2007 and their Associated Detectors CRANGE, TEST OR EXPERIMENT: FCR 78-031, which was completed on 5/15/78, changed the grounding on the radiation monitor chasis (in SEAS cabinets) from instru=ent to

- ctation ground. This change was recommended by the monitor vendct, Victoreen, and crrried out under a field change package prepared by the SFAS vendor, Consolidated C2ntrols. FCR 77-462, which was completed on 6/10/78, electrically isolated the d3tector canisters of the above radiation monitors (inside contain=ent); the detectors were grounded to station ground. By Victoreen's direction, they are new isolated from both station and instrument ground.

REASON FOR FCR: These FCRs were prepared to resolve an intermittent grounding problem which had been causing erroneous readings on radiation monitor RE-2006 (SFAS Channel 3). These changes were advised by Victoreen and Consolidated Controls in order to ground the monitors to station ground, as is all other Victoreen equipment at Davis-Besse Unit #1, and to eliminate undesirable ground loops.

SAFETY EVALUATION: These changes resolve the SEAS containment radiation monitor and detector grounding problems which should eliminate the erroneous readings and channel trips which have occurred in the past. Both the above changes were made et Victoreen's direction in order to improve the safety and reliability of their aquipment.

.S FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1978 FCR NO: 78-288 SYSTEM: Decay Heat Removal COMPONENT: Valve DH49, " Decay Heat Removal-Normal Cooldown Line Pressure Relief Check Valve" CHANGE, TEST, OR EXPERIMENT: FCR 78-288, which was completed on August 5, 1978, r;tated valve DH49 through a 45 degree angle. The valve was revelded into place, cud the necessary hydrostatic testing was completed.

REASON FOR FCR: The valve requires periodic maintenance to ensure that it does not 1sak by. The location was such that the bonnet could not be removed without removing the valve and associated piping from the decay heat valve pit. This change makes it psssible for the necessary periodic maintenance to be readily performed.

SAFETY EVALUATION: The rotation of this valve by 450 will not adversely change the function of check valve DH49. Rockwell, the valve manufacturer, agrees that this v;1ve will work properly with a 45 degree angle.

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s FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1978 FCR NO: 78-348 SY3 TEM: Component Cooling Water (CCW), Reactor Coolant Pump (RCP) Seals COMPONENT: Flow Indicating Switches (FIS) 4133, 4233, 4333, 4433 CHANCE. TEST, OR EXPERIMENT: On July 15, 1978, work was completed which removed the CCW low flow switches in the control circuits of the RCP seal return valves for each of the four reactor coolant pumps. This change was made under Maintenance Work Order (MWO) I&C 78-442 and a work package which was reviewed by the unit crchitect-engineer, Bechtel Corporation. The affected drawings are being revised by Bechtel.

REASON FOR THE FCR: This change was recommended by the nuclear steam supply system vendor, Babcock and Wilcox, upon investigation of the failure of two RCP seals.

The circuit was wired in such a manner that if low CCW flow occurs coincident with '

low seal injection water flow on a running RCP, the seal return valve would close.

Babcock and Wilcox recommended that the CCJ low flow switch contact be removed from the circuitry. Conversely, the valve will still close on an idle pump should loss of seal injection occur.

SAFETY EVALUATION: Removing the low CCW flow switches (CCW low flow interlock),

from the logic that closes the seal return valves (MU-59A, B, C or D) will remove cperation at off-design conditions. When first seal cavity CCW flow out from the RCP seal cooler drops below 45 gpa, (momentarily or for short periods of time due to changes in CCW flow within containment), FIS '*133 (4233, 4333, 4433) will no Icager close the seal return valve. Low seal injection flow will, however, still close the seal return valves. This change will enrance the safety function of the Reactor Coolant Pumps.

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