ML19209C207

Monthly Operating Rept for Sept 1979

ML19209C207
Person / Time
Site:	Davis Besse
Issue date:	10/08/1979
From:	Berger C, Stotz J TOLEDO EDISON CO.
To:
Shared Package
ML19209C204	List: ML19209C203 ML19209C207
References
NUDOCS 7910120202
Download: ML19209C207 (9)

Text

, .

AVERAGE DAILY UNIT PO'+ER LEVEL DOCKET NO.

50-346 gg7 Davis-Besse Unit 1 October 8, 1979 DATE

- d*" 3" '* ##1 *# 8 *'

COMPLETED BY 419-259-5000, Ext.

TELEPHONE 243 MONTil September, 1979 DAY AVERAGE DAILY POWER ' '_ VEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

, 877 880-37 2 869 18 470 3 860 19 477 4 . 877 20 862 5 871 21 885 6 876 - 22 883 7 785 3 848 8 0 24 881 9 0 25 879

. 10 174 26 767 11 441 27 0 12 423 28 131 13 655 29 '02 34 826 30 376 IS 882 3g 16 882 IN!iTRUCTIONS ,

On this format !Lr the sserage daily unit power level in MWe Net for each day in the reporting inonth. Compute to the nearest whole megswatt.

s (9/77) 1144 036 7910120 Mao

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OPERATING DATA REPORT DOCKET NO. oc 50-346 to od7, 1979 DATE COMPLETED BY J. Stotz/C. Berger TELEPHONE 419-250-5000, Ext.

242 OPERATING STATUS Notes

1. Unit Name: Davis-Besse Unit i Seo tember , 1979

2. Reporting Period:

2772

3. Licensed Thermal Power aMwt):

4. Nameplate Rating (Gross 31Wel:

925 906

5. Design Electrical Rating (Net MWe):

6. Ma.ximum Dependable Capacity (Gross MWe): to be detern;ined

7. Maximum Dependable Capacity (Net MWe):

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

9. Power Level To which Restricted. !f Any (Net MWe):

10. Reasons For Restrictions. !f Any:

"Ihis Month Vr to-Date Cumulative 720 6,551 18,316

11. Hours in Reporting Period 634.95 3,624.4 10,256.2

12. Number Of Hours Resetor Was Critical 85.05 1,943.3 2,733.6

13. Reactor Resene ShutdowTt Hours

14. Hours Generator On.Line 605.1 3,503 9,236.2 0 1,728.2 1,728.2

15. Unit Resene Shutdown Hours 1,549,866 8,681,528 16,869,098 .

16. Gross Thermal Energy Generated (MWH) 503,791 2,881,188 6,264,943

17. Gross Electrical Energy Generated (MhH) _

477,780 2,714,007 5,755,467

18. Net Electrical Energy Generated (MWH)

19. Unit Service Factor 83.9 55.5 50.4

20. Unit Asailability Factor 83.9 70.9 59.9

21. Unit Capacity Factor tU>ing MDC Net) to be determined

22. Unit Capacity Factor (Us ; DER Net) 73.2 J5. 7 34.7

23. Unit Forced Outage ".are 8.8 .9 19.0

24. Shutdowns Scheduled Oser Nest 6 Months (Type.Date.and Duration of Each1:

Refueline outace to start March 15, 1980

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

26. Units in Test Status t Prior to Commercial Operationi: Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY COMMERCI AL OPERATION

. 1144 037

t '

DOCKET No. '.0-346 UNIT SilU1 DOWNS AND POWut REDUCllONS UNIT N AMi, .31L iS-BeSRe UQit 1 ,

D ATr

• Octnhor R_ 1979 COMPLET ED liv Jan Stot7/Cqrl Berger REPORT hlONTil September, 1979 TEl.l.PIIONE _419-259-50' a; Ext. 243 I

"t. - '

Cause & Corree -

q .!E 'b .Y Licensec ,E3, '.,

Actiori in N". DJte R ~@

n s .2 s & Event e8 '

H $ 5 if, c Repori a N u? E' Prevent Recurrenes

$ e g u O

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NA NA Flaintenance outage to find st< ni ledk 11 79 09 07 S 58.6 B 1 NA in containment. .

NA NA NA Sticking pump prec'--- c'ontroller on l' 2 79 09 18 F 1/.2 A 3

'No. 2 electro-hydraulic control pump.

Controller to be disassembled and ,

i cleaned at earliest convenience.

13 79 09 23 S 0.0 B 4 NA NA NA Reactor power reduced to 75% to allow maintenance to adjust blowdown on SPl7A2.

41.1 NA NA NA Faulty capacitor on turbine throttle 14 79 09 26 F A 3 pressure transmitter power supply.

- 3 4

.I 2 Exhibit G. Instructions F: Forced Reason: klethod:

1.hf anual for Preparation of L'ata S: Schedu!cd A.Eiluipment Failure (Explain) 2 Manual Scram. Entry Sheets for Liecnsee B ht.tintenance oiTest Event Iteport (LElt) File (NU'IEG-

1. ' O. Refueling 3- Antonutic Scram.

1. 4-Other (E splain) 0161)

D Itegulatory itestris. tion I? Operator Training 1 License Examination C $

ILAdnunntiaisve

• l'd G Occiational I nor (limplain) Exhibit I Same Sourec Q (9/77) 11 Oihei (I spl.mi) ,

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OPERATIOMI,

SUMMARY

FOR SEPTEMBER,1979 -

'9/1/79 - 9/2/79 The reactor power level was maintained between 99 and 100 per-cent with the generator gross load at 915 i 10 MRe.

9/3/79 (Labor Day) Per Load Dispatcher, began reducing power at 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br /> to 10 percent. Began increasing power at 0850 hours0.00984 days <br />0.236 hours <br />0.00141 weeks <br />3.23425e-4 months <br /> and reached 99 oercent at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />.

9/4/79 - 9/6/79 The reactor power was maintained between 99 and 100 percent with generator gross' load at 915 1 10 MWe.

9/7/79 - 9/14/79 Maintenance outage to find source of steam leak in containment.

Reactor subcritical by 0220 hours0.00255 days <br />0.0611 hours <br />3.637566e-4 weeks <br />8.371e-5 months <br /> on Septe=ber 8, 1979. Returned to critical at 2105 hours0.0244 days <br />0.585 hours <br />0.00348 weeks <br />8.009525e-4 months <br /> on September 9,1979 and in Mode 1 at 2317 hours0.0268 days <br />0.644 hours <br />0.00383 weeks <br />8.816185e-4 months <br />. Power was increased to 50% on September 10, 1979 and held until September 13, 1979. Power reached 99 percent on September 14, 1979.

9/15/79 - 9/17/79 Reactor power level maintained between 99 and 100 percent with generation gross load at 915 1 10 MWe.

9/18/79 - 9/19/79 Turbine trip followed by reactor trip at 1243 hours0.0144 days <br />0.345 hours <br />0.00206 weeks <br />4.729615e-4 months <br /> on September 18, 1979 during electro-hydraulic coatrol pump ten _. A drop in electro-hydraulic control hydraulic pressure occurred when the pumps were switched over. The Anticipatory Reactor Trip S; stem tripped the reactor. Began recovery and increased peaar tc 99 percent by 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> on September 19, 1979.

9/20/79 - 9/22/79 The reactor power level was maintained between 99 and 100 percent with generation gross load at 915 i 10 MUe.

9/23/79 Reactor power reduced to 75 percent at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> to allow mainten-ance to reset the blowdown on SPl7A2. Returned to 99 percent power by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.

9/24/79 - 9/25/79 The reactor power level was maintained between 99 and 100 percent with generator gross load at 915 i 10 MWe.

9/26/79 - 9/29/79 Turbine / reactor trip at 2056 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.82308e-4 months <br /> on September 26, 1979 caused by a turbine throttle pressure transmitter power supply causing the turbine valves to partially close. Began increasing power on Septem-ber 28, 1979 and returned to 99 percent power on September 29, 1979.

1144 039

OPERATIONAL

SUMMARY

FOR SEPIEMBER,1979 PAGE 2 0F 2 9/30/79 The reactor power level was maintained between 99 and 100 percent with generator gross load at 915 i 10 MWe.

1144 040

e RET 1JELING INFOPJd.ATIC'l DATE: September, 1979

{

1, Name of facility: Davis-Besse Nuclear Power Station Unit 1 Seneduled date for next refueling shutdown: March, 1980 2.

3. Scheduled date for restart following refueling: May, 1980

4. Will refueling or rese=ption of operation thereaf ter require a technical

• If answer is yes, what, specificatica change or other license a=end=ent?

in general, will these be? If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Cc mittee

- to deter =ine whether any unreviewed safety questions are as=ociated with the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached .

M

5. Scheduled date(s) for submitting proposed licensing action and supporting infor=ation. December. 1979 Important licensing considerations associated with refueling, e.g., new or

( -

6.

different fuel design or supplier, unreviewed design or perfor:ance analysis methods, significant changes in fuel design,.new operating procedures.

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

~

7. The nc=ber of fuel asse=blies (a) in the core and (b)' in the spent fuel storage pool.

177 0 ( ero)

(a) (b)_

8. The pa esent 31 censed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned, e in nu=ber of fuel asse=$ lies.

260 Increase si:e by 475 (735 total)

Present

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity. .,

(.

Date 1989 (as.suming ability to unload the entire core into the spent fuel pool is =aintained and the unit goes to an 18 month refueling

) 1144 041

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. . i s

REEUELEIG INFOPJ#.ATION (Continued) .

Septeciber, 1979

• Page 2 of 2

4. The following Technical Specifications (Part A) vill require revision:

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Baes)

• 2.2.1 - Reactor Protection System Instrunente. tion Setpoints (and Bases) 3.1.3.6 - Regulating Rod Insertion Lf.=its

. 3.1.3.7 - Rod Program' 3.2.1 - Axial Power I= balance (and Bases)

The follovir.g Technical Specifications (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 Para =eters (and Bases) ,

9.

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-)

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. 1144 042 ,

FACILITY CHANGE REOUESTS COMPLETED DURING SEPTEMBER, 1979 FCR NO: 79-188 SYSTEM: Auxiliary Feedwater System COMPONE'!T: Auxiliary feedwater flow instrumentation CRANGE, TEST, OR EXPERIMENT: On June 28, 1979, the physical work for FCR 79-188 was completed. This FCR installed instrumentation to provide indication in the control room of auxiliary feedwater flow to each steam generator. This addition was made under the guidance of the unit architect / engineer, Bechtel Company. All affected drawings were revised by Bechtel to reflect the addition of this instrumentation.

REASON FOR THE FCR: This instrumentation was installed in order to comply with a ce=nitment documented in the May 16, 1979 Shutdown Order.

SAFETY EVALUATION: The addition of the auxiliary feedwater flow indication does not affect the function of the auxiliary feedwater system in any way. Ultrasonic flow transmitters were utilized to sense the flowrate through the auxiliary feedwater 01 ping to each steam gener, tor. The flow transmitters are clamped to ;he outside of the piping and do not penetrate the pressure boundary. It has been verified by the architect / engineer that the small additional weight of the transmitters does not affect the seismic qualification of the piping.

FACILITY CFANCE REQUESTS COMPLETED DURING SEPTDiBER, 1979 FCR NO: 79-251 SYSTEM: Decay Heat Fessval COMPONE';T: Hydraulic snubber CCA-4-H3 CHANGE, TEST, OR EXPERIMENT _: On June 27, 1979, under Maintenance Work Order 79-2282, hydraulic snubber CCA-4-H3, which was a remote reservoir snubber, was replaced with a local reservoir snubber.

REASON FOR THE FCR: The change to a local reservoir snubber allows for easier re-

= oval and replacement of the snubber. It was also found that the original snubber had non-ethylene propylene seal materials (see Licensee Event Report NP-32-79-07).

SAFETY EVALUATION: The hydraulic shock and sway suppressor being installed is the type called for on existing project drawings. The new snubber has ethylene propylene seals and is in accordance with technical specifications. The local reservoir location and new snubber will not a lver sely af f ect the safety function of this support. This is not an unreviewed safety question.

O e

i 144 044