ML20148H741

Monthly Operating Rept for Oct 1978

ML20148H741
Person / Time
Site:	Davis Besse
Issue date:	11/08/1978
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML20148H739	List: ML20148H734 ML20148H741
References
NUDOCS 7811140136
Download: ML20148H741 (9)

Text

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O OPER ATING DATA REPORT I

50-346_

DOCKET DATENO.11/DI # D COMPLETED itY Ercal 1 M aba i TELEPil0NE CfF25'975000, Ex t . ,

236-OPERATING STATUS

1. Unit Name: Davis-Besse Unit 1 Notes

2. Reporting Period: October, 1978

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

7. Maximum Dependable Capacity (Net MWe): to be determined

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

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9. Power Level To Which Restricted,if Any(Net MWe): None

. 10. Reasons For Restrictions,If Any:

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6 This Month Yr -to.Date Cumulatise 4

11. Ilours In Reporting Period 745 7296 10301 2

12. Number Of Hours Reactor Was Critical 265.4 3819 5611.1 1 l

13. Reactor Resene Shutdown flours 0 38.9 422.6

14. Ilours Generator On.Line 138.5 3384.9 4851.7

. 15. Unit Resene Shutdown Hours 0 0 0

16. Gross Thennal Energy Generated (MWil) 341.841 6,516,638 8,180,670 2,733,720 17 Gross Electrical Energy Generated (MWH) 109,873 _ _ 2,208,771 a 18. Net Electrical Energy Generated (MWH) 92.613 2,014,752 2,444,570

19. Unit Senice Factor 18.6% 46.4% 49.9%

< 20. Unit Asailability Factor 18.6% 46.4% 49.9%

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

22. Unit Capacity Factor (Using DER Net) 13.7% 30.5% ~~3I~74

23. Unit Forced Outage Rate 29% 25.8% 25.2%

24. Shutdowns Scheduled Oser Next 6 Months (Type. Dale, and Duration of Each):

November 10, 1978

25. If Shut Down At End Of Report Pen. d. o Estimated Date of Startup-

{ 26. Units in Test Status (Prior to Commercial Operation): Foiecast Achiesed INITI A L CRITICA LITY 8/12/77

. INITI A L ELECTRICITY 8/28/77 CON!MERCIA L OPER ATION 11/M//7*

• Declared operational at 25% 12/19/77**

• • Declared operational at 40% (from 25%) 1/23/78***

• • • Declared operational at 75% (from 40% 7/31/78****

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• • • • Declared operational at 100% (from 75 )

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l 4 . AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-346 UNIT Davis-Besse Unit #1 11/8/78

DATE CO\1PLETED BY Erdal C. Caba

' TELEPilONE 419-259-5000, Ext.

236 i

October, 1978 MONTil i

l DAY AVER AGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL 3

(M We-Net ) (MWe-Net )

1 0 g7 0 2 11 Ig 0 213 0 4, 3 19

- 595 0 4 20 885 0 1 5 ,

838 0 4 6 22  ;

7 828 0 23 i

834 0 8

42 0 9 25 0

l 10 26 0 0 11 27

' 0 12 28 0 0 13 79  ;

14 0 30 0 0 0 15 3, 16 0 l

\l INSTRUCTIONS On this format, list the average daily unit power leselin MWe. Net for each day in the reporting month. Coinpute to the nearest whole megawatt.

01/77) 1

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DOEKET NO. 50-346 1& Nil SilUIDOWNS AND POKM REDUCllONS' '

UNIT NAM E Davis-Besse Unit 1 DATE 11/6//6 REPORT MONTil October, 1978 COMPLETED !!Y Charles N. Alm TELEPIIONE 419-259-5000, Ext. 251

, .5 ? 3 Y Licensec ,E g _fvt Cause & Corrective No. Date o. 3g ( ,3 5 5 Event g o-3 Action to

$~ $ Report a v) U 8O Prevent Recurrence 5

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24 continued F 43.2 A 3 NA IA INSTRU The reactor tripped because of rea'c-from last tor coolant system (RCS) low pressure.

month The RCS low pressure occurred when the steam generators were overfdd dur-ing an atterapt to stabilize plant para -

meters. The transient parameters were ,

' caused by the failure of the RCS flow transmitter for Loop 2 of the RCS.

25 78 10 03 F 13.4 A 3 NA IIA (Unknown) Refer to operational summary on Octo-ber 3, 1978.

26a 7'u 10 09 S 432.5 A 1 NA CB PUMPXX The unit was shutdown to replace the seals of Reactor Coolant Pumps 2-1 and 2-2. Refer to the operational summary for details on outage.

3 4 I 2 F: Forced Reason: MethoJ: Exhibit G-Instructions A Equipment Failure (Explain) 1 -Manual for Preparation of Data S: Schedu!ed B-Maintenance of Test 2 Alanual Scra m. Entry Sheets for Licensee C Refucting 3-Automatic Scram. Event Report (LER) File (NUREG-D Regulatory Restriction 4-Other ( Explain) 0161)

E-Operator Training & Liccuse Examination

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F-Adniinistiative ~ 5 G-Operational Er ror (lixplain) Exhibit I . Same Source (9/77) Il-Ot her ( E xplain)

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DOCKET NO.

50-346 .

UNil SIIUIDOWNS AND PON.;l REDUC 110NS '

UNIT N AME Davis-Besse Unit 1 +

DATE 11/8/78 -

COMPLETED BY rhnri n e: y_ 33, REPORT MONT,- October. 1978 TELEPilONE 419-259-5000, Ext. 251 l

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-, jg 3 $ E2 Licensee ,E t, ""1, Cause & Corrective No. Date c.  ;= @ js& Event a7 g 9 'd Action to C j@ d Report n AG jU Prevent Recurrence j ;f, g g ..

116.4 NA NA NA N The unit was still shutdown, however, 26b, 78 10 27 S B the reactor was critical and maintainet at 4 percent power in preparation for the Natural Circulat, ion Flow Test.

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! I F: Forced iteamn: Method: Exhibit G Instructions A lh ioipment Failure (Explain) 1 -Manual for Preparation of Data l S: Schedu!ed l D-Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee c-Itcfueling 3-Automatic Scram. Event Report (LER) File (NUREG-l D Itegulatory itestriction 4-Other (Explain) 0161)

E-Operator Training & License Examination F- Att min e.t ra tive 5

! G. Operational Eirur (Explain) Exhibit I - Same Source (9/77) Il-Ot her (Explain) l l

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OPERATIONAL

SUMMARY

FOR OCTOBER 1978 4

10/1/78 The repair work on the Reactor Coolant System (RCS) Loop 2 flowmeter was completed, and unit startup was initiated.

The repair work on the flow transmitter required amplifier board replacement and calibration.

4 10/2/78 Reactor criticality was attained at 0641 hours0.00742 days <br />0.178 hours <br />0.00106 weeks <br />2.439005e-4 months <br />, and power escalation was initiated. The turbine-generator was syn-chronized on line at 1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br /> with the reactor at 15 per-cent power.

10/3/78 The turbine-generator tripped at 1207 hours0.014 days <br />0.335 hours <br />0.002 weeks <br />4.592635e-4 months <br /> which initiated a reactor trip on low pressure at 1208 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.59644e-4 months <br /> during the run-back. Reactor power had just attained 75 percent power with the generator gross load at 675 MWe when the turbine trip oc cu rr ed .. The turbine trip was initiated by low electro-hydraulic control (EHC) pressure when the second EHC pump was started. Investigation for the cause of the EHC pressure transient was initiated, but no equipment malfunctions were l discovered. During the next unit shutdown, the EHC pump dis-  !

charge check valves and relief valves will be inspected and  !

further testing with strip chart recorders for pressure tran-sients will be initiated.

I Reactor criticality was re-established at 1928 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.33604e-4 months <br />.

10/4/78 The turbine-generator was synchronized on line at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br />, and power escalation continued. The unit power output attained 100 percent full power at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> with the genera-tor gross load at 930 + 10 MWe.

1 10/5/78 - 10/8/78 The unit power level was maintained at 94 percent power during this time span.

10/9/78 - 10/26/78 The unit was shutdown to replace the seals on Reactor Ccolant Pumps 2-1 and 2-2. The shutdown was initiated at 0048 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> on October 9, 1978.

In conjunction with the seal replacement work, the following repair work was completed:

(1) Three leaking condenser tubes were located and plugged.

Also, the condenser was staked per Facility Change Request 78-480 to reduce tube failure due to vibration.

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, OPERATIONAL SUMSIARY FOR OCTOBER 1978 PACE 2 i

l (2) The thrust bearings on the Main Feedwater Pumps were adjusted to reduce vibration.

(3) Casing leaks on the Booster Feedwater Pumps were l

^ repaired.

(4) Seals on the condensate polisher valves were replaced to prruent further operational problems.

' (5) The steam generators were drained and bonnet leaks on instrumentation isolation valves were repaired.

(b) The EHC check and relief valves were inspected to deter-

mine the cause of the pressure transient on October 3, 1978. A nick in the seat of one check valve was dis-covered and repaired. l (7) Casket f ailure on the Moisture Separator Reheater 1-2 was repaired.

The outage duration was increased on October 15, 1978, when 1 the auxiliary impeller of Reactor Coolant Pump 2-1 was dis- I covered damaged and had to be machined. The damage was I caused by a seal cartridge bolt which had backed out from the I bottom of the seal cartridge when the locking wire broke. The bolt was found lying in the bottom of the seal cavity on top of the auxiliary impeller.

On October 19, 1976, the outage completion was delayed be-cause the RCS had to be drained again because of no seal injection flow to Reactor Coolant Pump 2-2. The line was discovered plugged by crystalized boric acid.

10/27/78 Reactor criticality was attained at 0336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> and the reactor i l

power was maintained at 4 percent power to enable xenon equili-brium in preparation to perform the Natural Circulation Flow Test (TP 800.04).

10/28/78 The Natural Circulation Test (TP 800.04) performance was initiated at 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br />.

10/29/78 - 10/31/78 The reactor tripped because of RCS low pressure at 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br /> on October 29, 1978. The trip investigation revealed that the low pressure was caused by either the electromatic relief valve or spray valve on the pressurizer failing to close at

. w OPERATIONAL SIRDIARY FOR OCTOBER 1978 PACE 3 the proper time after actuation. Proper response of these valves was required because of a series of Steam and Feed-water Rupture Control System (SFRCS) trips. The series of SFRCS trips were initiated by a malfunction of the speed changer on the Main Feedwater Pump Turbine 1-2.

i Reactor criticality was re-established at 2012 hours0.0233 days <br />0.559 hours <br />0.00333 weeks <br />7.65566e-4 months <br /> on October 29, 1978, and reactor power was maintained at 4 percent power the remainder of the month to perform the Natural Circulation Flow Test. On October 30, 1978, the Natural Circulation Flow Test was delayed for the remainder of the month because of the inability to meet B&W chemistry specification for dissolved oxygen in the Condensate Storage Tank.

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g j FACILITY CHANGE REQUESTS COMPLETED DURING OCTOBER, 1978 1

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FCR NO.77-424 l SYSTEM

Steam and F.eedwater Rupture Control System (SFRCS)

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i COMPONENT: Pressure Differential Switches (PDS) 2685A thru D, and PDS 2686A thru D f

CHANGE, TEST, OR EXPERIMENT: On September 28, 1978, work was completed which

{ relocated the high pressure tap connections for PDS 2685A thru D and PDS 2686 i A thru D as per FCR 77-424. The function of these pressure switches is to j detect a main feedwater line rupture and actuate the auxiliary feedwater sys-l tem via the SFRCS. The change made was to move the high pressure tap connec-j tion for each switch from the auxiliary f eedwater line downstream of the check valve to downstream of the main feedwater check valve. The switches now j detect the pressure diff erential across the main f eedwater line check valves.

I 3 REASON FOR THE FCR: The former connection of the pressure switches, between l the main feedwater line and the auxiliary feedwater line, was causing spurious

trips of the SFRCS and unnecessary unit shutdowns. The spurious trips were

being caused by perturbations in main feedwater line pressure.

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SAFETY EVALUATION: The following analyses are to describe why the tap reloca-

' tions will not affect the ability of the system to detect e main f eedwater line rupture.

j 1. A main feedwater line rupture upstream of the main feedwater line check j valve: A main feedwater line pressure will drop and the check valve essu e acro s h cleck val o in re se I he le rsueo the steam generator side of the check valve exceeds the pressure on the main feedwater side of the check valve by 197.6 psi, the auxiliary feed- i water system will be initiated. This will be caused by the actuation of i the differential pressure switch. l

2. A main feedwater line rupture downstream of the main f eedwater line check valve: The differential pressure switches will not detect this f ailure mode. 'During this failure mode the water level in the steam generator will drop. The steam generator low level instrumentation will detect this rupture and will initiate the auxiliary feedwater systen.

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FACILITY CilANGE REQUEST 77-424 PACE 2 0F 2

3. A main f eedwater line rupture upstream of the main f eedwater line check valve, with a main feedwater line check valve f ailure: The differen-tial pressure switches will not detect this failure mode. During tb ts f ailure mode, the differential pressure across the open check vel - uill not cause the differential pressure switches to actuate. The water level in the steam generator will drop. The steam generator low level ins:ru-mentation will detect this rupture and will initiate the auxiliary feed-water system.

This change will not adversely affect the safety function of the system.

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