ML19261B307

Monthly Operating Rept for Jan 1979

ML19261B307
Person / Time
Site:	Davis Besse
Issue date:	02/09/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19261B302	List: ML19261B301 ML19261B307
References
NUDOCS 7902150295
Download: ML19261B307 (14)

Text

.

AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

UN17 Davis-Besse Unit 1 February 9, 1979 DATE CO3fPLETED BY Erdal C. Caba 419-259-5000, Ext.

TELEPilONE 236 January, 1979 Af0NTil DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (&f We-Net )

381 g7 0 1

791 gg 0 2

881 0 3 39 874 0-4 .

20 863 0 5 21 876 0 6 22 876 0 7 23 875 0 8 .24 875 0 9 25 671 0 10 26 726 27 11 284 0 12 28 238 17 13 29 00 30 14 0 729 3g 15 16 0

INSTRUCTIONS On this format, list the average daily unit power leselin MWe-Net for each day in the reporting month. Compute to the riearest whole megawatt. ,

(9/77)

• 4 vsozisoMf

OPERATING DATA REPORT DOCKET NO.TLbruary 50-346 9,1979 DATE COMPLETED BY Erdal C. Caba TELEPil0NE 419-259-5000, Ext.

236 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

2. Reporting Period: January, 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

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

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): None

10. Reasons For Restrictions,If Any:

This Month Yr. to Date Cumulative 744 744 12,509

11. Ilours In Reporting Period

12. Number Of flours Reactor Was Critical 381.3 381.3 7,013.1 0.8 R.5 4u.4

13. Reactor Reserve Shutdown llours

14. Ilours Generator On-Line 346.7 346.7 6.079.9 0 0 0

15. Unit Reserve Shutdown Hours

16. Gross Thermal Energy Generated (MWH) 828.938 828.938 11,016,508 .

17. Gross Electrical Energy Generated (MWH) 274.288 _ _, 274.288 3,658,043

18. Net Electrical Energy Generated (MWH) 252.823 252.823 3.294,283 46.6% 46.6% 51.1%

19. Unit Senice Factor

20. Unit Availability Factor 46.6% 46.6% 51.1%

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

22. Unit Capacity Factor (Using DER Net) 37.5% 37.5% 34.7%

23. Unit Forced Outage Rate 7.7% 7.7% 25.9%

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

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

26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved INITIA L CRITICA LITY INITI A L ELECTRICITY COMMERCI AL OPER ATION (9/77)

t DOCKET NO. _ 50-346 -

UNIT SilUTDOWNS AND PObUt REDUCTIONS UNIT N AME Davis-Besse Unit 1 DATE Charles N. Alm COMPLETED BY Januarv. 1979 TELEPil0NE 419-2s9-5000 Ext. 251 REPORT MONTil e_.

Cause & Corrective ,

jg 3 Y Licensee Eg '. Action to [-

ss Event p g-No. Dale g H

3g g:

s

$ j;g Report a mO }g Prevent Recurrence 6 .

The unit was shutdown to repair the 1 N/A Illi PIPEXX 32 78-12-16 F 3.7 A extraction steam line bellows. Refer -

(Cont' d) ,

to the Operational Summary of Decom-ber for further details ,

Power was reduced to 75 percent to 4 N/A N/A N/A 0.0 B

~

33 79-01-10 S perform a Power Imbalance Detector *

' Correlation Test. .

EB ELECON The event was initiated by the acci-F 25.2 3 NP-33-79-13 dental grounding of the liydrogen 34 79-01-12 11 Analyzer AE 5028. The ground caused the 200 amp fuse on the inverter feed-ing Y2 Essential 120 VAC Instrument Bus to blow. The 10 amp fuse on the

. Ilydrogen Analyzer AE 5028 was a standard fuse. The fuse was replaced with a quick acting fuse. .

4 3

I 2 Method: Exhibit G. Instructions Reason: for Preparation of Data F: Forced 1-Manual S: Schedu!ed, A Equipment Failure (Explain) Entry Sheets for Licensee 2 Manual Scram. Event Report (t.ER) File (NUREG-B. Maintenance of Test 3 Automatic Scram.

C Refueling 4-Other (Explain) 0161)  !

D-Itegulatory itest riction  ;

E-Operator Training & License Examination 5 ~

F Adminntrative Exhibit I- Same Source ,

G. Operational Ee ror (Explain) t Il Osher (Explain) 19/77)

-- .h

t  ;

DOCKET NO.

50-346 .

UNIT SilUIDOWNS AND POWEll REDUCTIONS UNIT NAMEFebruary Davi s-Besse I' nit 1

9. 1979 DATE Charles N. Alm COMPLET ED llY REPORT MONTil Januarv_ 1979 TELEPilONF 419-259-5000 Ext. 251 w

t.

= c

• Cause & Corrective ,

,$g 3 Licensee @,, .,

i No. Date g 3g i!i 3.Ei ss Event gg l"1

g. V Action to Prevent Recurrence mO g F

f5 5 j;gg Report #

6 4 N/A N/A N/A The turbine was tripped from 100 per-35 79-01-14 S .9 B cent power to ccxnplete the Unit Load

- Rejection Test, TP 800.13. .

1 4 N/A N/A N/A The reactor was shutdown twice in 36 79-01-14 S 359.5 B this time frame to complete the unit +

tests TP 800.25, " Shutdown From Out-side the Control Room", and TP 800.26,

., " Loss of External Load Including Loss '

of Of f site Power."

~. ,

A unit outage was then continued af ter the shutdown of TP 800.26 to repair many equipment f ailures. The major failure which required the outage was

, to replace the seals on Reactor doolant Pumps 1-1 and 2-1. Refer to the Operational Summary for further .

d e'ta il s .

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

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

D. Regulatory Restriction E Operator Training & License Examination 5

• F-Administ rative Exhibit I Same Source G-Operational Litor (Explain) 08/17)

Il-Other (Explaini

OPERATIONAL

SUMMARY

FOR JANUARY, 1979 1/1/79 The turbine-generator was synchronized on line at 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br />.

Reactor power was then slowly increased and attained 83 per-cent at 2208 hours0.0256 days <br />0.613 hours <br />0.00365 weeks <br />8.40144e-4 months <br /> with the generator gross load at 790 i 10 MWe.

J/2/79 Reactor power was _ increased to 91 percent at 1042 hours0.0121 days <br />0.289 hours <br />0.00172 weeks <br />3.96481e-4 months <br /> with the generator gross load at 860 1 10 MWe. The power level had been maintained at 83 percent since Turbine Control Valve Number 3 oscillated et higher power and consultation with General Electric was therefore necessary prior to increasing power.

1/3/79 Reactor power attained 100 percent at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> af ter suffi-cient soak time at 91 percent was provided to obtain required xenon equilibrium. The generator gross load at 100 percent was 925 1 10 MWe.

1/4/79 Reactor power was decreased to 65 percent at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> to l perform turbine stop valve testing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to load ,

i rejection testing.

l 1/5/79 The turbine stop valve testing was completed at 0028 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />, and reactor power was then increased to perform turbine con- ,

. trol valve testing. The control valve testing was completed at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, and reactor power was returned to 100 percent at 0548 hours0.00634 days <br />0.152 hours <br />9.060847e-4 weeks <br />2.08514e-4 months <br />.

1/6/79 - 1/9/79 Reactor power was maintained at 100 percent until 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> on January 9,1979 when a power reduction to 75 percent was initiated. The load rejection testing was delayed because of possible Reactor Coolant Pump (RCP) seal degradation.

1/10/79 Reactor power was maintained at 75 percent to enable xenon equilibrium prior to performing the Power Lnbalance Detector Correlation Test. The generator gross load at this power level ves 715 i 10 MWe.

1/11/79 The Power Imbalance Detector Correlation Test was completed at 1346 hours0.0156 days <br />0.374 hours <br />0.00223 weeks <br />5.12153e-4 months <br />, and reactor power was increased at 1508 hours0.0175 days <br />0.419 hours <br />0.00249 weeks <br />5.73794e-4 months <br />.

Reactor power attained 100 percent at 2135 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.123675e-4 months <br />.

l PAGE 2 r OPERATIONAL

SUMMARY

FOR JANUARY, 1979 1/12/79 Reactor power was decreased to 65 percent at 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to enable turbine stop valve testing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the per-formance of the 100 percent power Load Rejection Test. Reac-tor power had been returned to 100 percent at 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br />. l.

A reactor /turbi 2 trip occurred at 0933 hours0.0108 days <br />0.259 hours <br />0.00154 weeks <br />3.550065e-4 months <br />. The trip followed a loss of Reactor Coolant System (RCS) flow indica-tion to the Integrated Control System (ICS) and a partial loss of neutron power indication to the ICS. The event was initiated by an accidental grounding of the Containment Hydrogen Analyzer AE 5028. The ground was large enough to blow the 200 amp inter-nal inverter fuse on the inverter feeding Y2 Essential 120 VAC Instrument Bus. The loss of the Y2 Bus was investigated and corrected as per the Licensee Event Report NP-33-79-13.

Reactor criticality was re-established at 0446 hours0.00516 days <br />0.124 hours <br />7.374339e-4 weeks <br />1.69703e-4 months <br />, and the 1/13/79 turbine-generator was synchronized on line at 1044 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.97242e-4 months <br />. At 1815 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.906075e-4 months <br />, reactor power was held at 62 percent for thirty minutes to complete the turbine stop valve testing and was then increased to 92 percent. At 2348 hours0.0272 days <br />0.652 hours <br />0.00388 weeks <br />8.93414e-4 months <br />, reactor power was in-creased to 100 percent power.

The turbine was tripped per TP 800.13, " Unit Load Rej ection 1/14/79 Test" at 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The reactor did not trip, andAtthe turbine-0242 hours, generator was synchronized on line at 0052 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br />.

recctor power was leveled off at 41 percent with the turbine-generator gross load at 260 + 10 MWe.

A reactor power decrease to 15 percent was then initiated at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> in preparation to perform the test TP 800.25, " Shut-The reactor was tripped down From Outside the Control Room".

at 1015 hours0.0117 days <br />0.282 hours <br />0.00168 weeks <br />3.862075e-4 months <br /> per TP 800.25.

The reactor was returned to criticality at 2142 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.15031e-4 months <br />.

The reactor was shutdown at 1024 hours0.0119 days <br />0.284 hours <br />0.00169 weeks <br />3.89632e-4 months <br /> from 20 percent power to 1/15/79 perform the TP 800.26, " Loss of External Load Including Loss of Offsite Power".

A unit outage was initiated, and the following major work items 1/16/79 - 1/28/79 were completed:

PAGE 3 OPERATIONAL

SUMMARY

FOR JANUARY, 1979 I

1) The PCT l-1 and 2-1 seals were replaced.

2) The Number 8 turbine bearing was disassembled to repair an oil leak.

3) Further checkout of the Fuel Handling Bridges were made, and a new pin was installed in the main bridge mast. Also, the hydraulic hoses were identified as worn and will be replaced in a future outage.

4) The remaining sets of RCP capacitors were changed per West-inghouse's request to eliminate potential f ailures.

5) The pressurizer code safeties RC13A and RC13B and the elec-tromagnetic relief valve RC2A were disassembled, inspected, and repaired under directions of the Crosby vendor.

6) The Main Steam Isolation Valve (MSIV) pushbutton was re-located, and the Steam and Feedwater Rupture Control Sys-tem (SFRCS) half trip modifications were completed.

7) The pressurizer heater breakers were renlaced. All breakers were not changed though due to the inability to get spare breakers from the vendor because of w2ather conditions.

8) The generator high voltage bushing was inspected by vendor and replacement was determined necessary during next refuel-ing outage.

9) The Diesci Generators were inspected to determine cause of reverse power trips. Apparently the reverse power trip relays pick up too quick and further engineering evaluation must be made to resolve this problem.

10) The failure of Core Flood Tank Isolation Valve CFlA to oper-ate because of a cracked motor pinion gear was repaired.  ;'

11) The operational problems with the three way flow valve MU-11 was investigated and repaired. The repair involved pinning the coupler and actuator stem together to prevent slippage of the stem.

12) The Main Feedwater Pump 1-1 was disassembled, and new seals were installed to reduce excessive seal leakage.

OPERATIONAL SUSDfARY FOR JANUARY, 1979 PAGE 4 .,

13) Both Booster Feedwater Pumps were disassembled to replace casing seal gasket with a new type to reduce casing leaks.

14) The control rod position indicator for Group 6 Rod 3 was replaced.

15) Fcur hotleg thenmowells were seal welded af ter observation of leakage.

16) Two 18" extraction steam line bellows (expansion points) {

were replaced. One of these bellows was replaced last Dec ember . During Ecolaire evaluation of previous failures, the design problem was identified and new expansion joints were supplied and installed.

17) The o-ring seals on selected incore tubes were replaced to reduce excessive leakage.

18) High pressure feedwater heater 1-4 modifications were initiated and were still in progress at the end of the month.

1/29/79 The reactor attained criticality at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />, and the turbine-generator was synchronized on line at 1748 hours0.0202 days <br />0.486 hours <br />0.00289 weeks <br />6.65114e-4 months <br />.

1/30/79 At 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br />, the Main Steam Safety Valves SPl7A2 and SPl7B2 were declared inoperable and reactor power must remain below 90 percent until the problem is resolved.

1/31/79 Reactor power was maintained at 87 percent with the generator gross load at 790 + 10 MWe. The feedwater heater train 1 was still isolated, and the Main Stean Safeties SPl7A2 and SPl7B2 were still inoperable.

FACILITY CHANGE REQUESTS COMPLETED DURING JANUARY, 1979 FCR NO: 77-072 SYSTEM: Emergency Ventilation System COMPONENT:

Annalus Pressure Differential Strings (PDT-5000 & PDT 5014)

CHANGE, TEST, OR EXPERIMENT:

On January 20, 1979, the work for FCR 77-072 was c omple ted . This FCR initially required the installation of a compound pressure A supplemental change to the FCR gauge with the range of + 2 inches of water.resulted in the installation of a vacuum water vacuum.

This change was documented and approved by the unit architect-engineer, Bechtel Corporation. I REASON FOR THE FCR:

The pressure indicator previously installed was not scaled to accurately indicate in the range required (0.25 to 1.25 inches of water vacuum).

SAFETY EVALUATION:

This change does not affect the safety of the system or unit.

It will further aid in the correct evaluation of surveillance tests.

\

FACILITY CHANGE REQUESTS COMPLETED DURING JANUARY,1979_

FCR NO: 77-101 SYSTEM: Containment Ventilation COMPONENT:

Containment Air Sample Isolation Valve (CV 50llB)

CHANGE, TEST, OR EXPERIMENT: On May 26, 1978, the work for FCR 77-101 was com-pleted. This FCR modified the seismic support clamp on the Containment Air The change was j Sample Isolation Valve CV 5011B to be adjustable in width. 6 designed and the af fected drawings revised by the unit architect-engineer, Bechtel Corporation. ,

The seismic support around the motor was so close to it that REASON FOR THE FCR: It would have been necessary to cut off the the motor could not be removed.

seismic support to replace or adjust the motor.

This change which revises the seismic support around the SAFETY EVALUATION:

Containme;.t Air Sample Isolation Valve CV 5011B motor, will permit motor adjust-The change will not adversely af fect the safety function of ments and removal.

the Containment Ventilation System because system integrity is preserved by proper redesign of the seismic support.

e

= .

Is FACILITY CHANGE REQUESTS COMPLETED DURING JANUARY,1979 FCR NO.77-154 SYSTEM: Main Steam Line Area COMPONENT: Pressure Doors 215, 601, 602 Revise Bechtel Specification 7749-A-2, Revision 5, CHANGE, TEST, OR EXPERIMENT:

Section 15.2, " Testing", to read:

" Hatch and doors shall be tested for leakage by pressurizing the spa between the double seals at i as specified in American Society for for Testing and Rate of Air Materials Leakage (ASTM) des g-Through nation E-283 Standard Method of Test The seals shall be adjusted Exterior Windows, Curtain Walls, and Doors.

to allow a maximum flow rate of 2.0 SCFH at a constant 10 psig pressure.

Af ter the seals have been adjusted, an acceptance test shall be witnessed by the BUYERS and CONSTRUCTION MANAGER'S inspectors to determine th reliability of the readings".

REASON FOR THE Fjy1:

The original test method which involved the use of a pressure The new method is consistent with the procedures decay method was not realistic.is the ; cepted practice, and it satisfied the basic described intent.

in ASTM E-283, itThis change was made by the unit architect-engineer, Bechte with the guidance of the manuf acturer, overly Manufacturing Company.

The basic function of the doors is not affecced by this change.

SAFETY EVALUATION:

The FCR replaces an unrealistic test method with a test method consistent with accepted practice as per ASTM E-283.

(FSAR).

This change has no ef fect on the Final Saf ety Analysis Report

FACILITY CHANGE REQUESTS COMPLETED DURING JANUARY,1979 FCR NO: 77-432 SYSTEM: Auxiliary Feedwater (AFW) System COMPONENT:

AFW Pumps (P14-1, P14-2) and Turbines (K3-1, K3-2)

This CHANGE, TEST, OR EXPERIMENT:

On January 3,1979, FCR 77-432 was completed.

FCR required the change of drawings and schedules to reflect the correct descrip-tion and design parameters for a portion of This the Auxiliary Feedno FCR required Pump, turbine, physical andto change cooling water lines (HBD-137 to HBD-210) .All drawings and schedules have been change the this system FCR. piping.This change was documented and approved by the unit architect-engineer, Bechtel Corporation.

REASON FOR THE FCR:

The previous piping class service number HBD-137 did not ade-quately describe the service and design conditions of the piping system.

SAFETY EVALUATION: This change will not affect the safety function of the system.

The system design was not altered.

FACILITY CHANGE REQUESTS COMPLETED DURING TANUARY,1979 FCR NO.77-464 SYSTEM: Nuclear Filter Cask Handling (Clean Waste Monitor Tank Filter F9-1, F9-2)

COMPONENT: Unistrut Conduit Hanger 4367-207-SE CHANGE, TEST, OR EXPERIMENT: On November 21, 1978, all physical work and testing pertaining to FCR 77-464 was completed. FCR 77-464 shortened the horizontal span of the Q-listed unistrut conduit hanger, number 4367-207-SE by 10 inches. This change was documented by and approved by the unit architect-engineer, Bechtel Corporation. The affected drawing E-302A Sheet 261, was revised to reflect the above change.

REASON FOR THE FCR: The hanger in question was longer than it needed to be and as a result impeded the proper positioning of the nuclear filter cask when replacing the nearby Clean Waste Monitor Tank filter.

SAFETY EVALUATION: This change will not affect the saf ety function of the system.

The Q-listed conduit will be adequately supported af ter the change is completed.

. Revision 1 - 2/9/79 OPERATING DATA REPORT DOCKET NO. 50-346 DATE January 6, 1979 COMPLEJED BY Erdal Caba TELEPHONE 419-259-3000, Ext.

236 OPERATING STATUS

1. Unit Name: Davis-Besse Unit 1 Notes December, 1978

2. Reporting Period:

77

3. Licensed Thermal Power (51Wt):

923

4. Nameplate Rating (Gross A1We):

906

5. Design Electrical Rating (Net MWe):

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

7. Maximum Dependable Capacity (Net MWe):

to be determined

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

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

10. Reasons For Restrictions,if Any:

This Month Yr.-to Date Cumulative 744 8760 11765

11. Hours In Reporting Period

-328.8 4839.7 M.o

12. Number Of Hours Reactor Was Critical

13. Reactor Reserve Shutdown Hours 0 38.9 422.6

14. Hours Generator On-Line 314.1 4266.2 3/33.4 0 0 0

15. Unit Reserve Shutdown Hours

16. Gross Thermal Energy Generated (MWH) 626,864 8,523,538 10,187,570 .

196.972 2,859,306 3,383,755

17. Gross Electrical Energy Generated (MWH) _,

18. Net Electrical Energy Generated (MWH) 173,312 2,611,642 3,041,460 42.2% 48.7% 51.4%

19. Unit Service Factor

20. Unit Availability Factor 42.2% 48.7% 51.4%

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

22. Unit Capacity Factor (Using DER Net) 25.7% 32.9% 33.3%

23. Unit Forced Outage Rate 57.8% 25.4% 26.7%

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

None

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

26. Units in Test Status (Prior to Commercial Operation): Forecast Achieved INITIA L CRITICA LITY INITIAL ELECTRICITY N/A COMM ERCI A L OPER ATION (4/77) i i