Text

Monthly Operating Rept for Oct 1983
	ML20082S956
Person / Time
Site:	Davis Besse
Issue date:	11/09/1983
From:	Murray T, Sarsour B; TOLEDO EDISON CO.
To:	Haller N; NRC
References
	K83-1583, NUDOCS 8312150082
	Download: ML20082S956 (27)
	v • d • e

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse #1 DATE Novetsber 9. 1983 COMPLETED B Bilal Sarsour TELEPHONE (419) 259-5000 ext. 384 MONTH October, 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe Net)

I 115 37 368 -

a 2 0 Ig 367 3 n 39 373 4 . 184 20 393 5 272 21 391 6 272 22 390 7 267 33 505 8 42 624 24 9 249 25 646 10 367 26 631 11 364 27 650 12 366 28 648 13 504 '

29 14 626 652 30 15 40 3, 785 16 .

216 i INSTRUCTIONS On this fortnat. list the average daily unit power level in MWe Net for each day in the reportmg month $ Compute to l the nearest whole megawatt. .

I"I77 3 8312150002 831109 PDR ADDCK 05000346 R PDR

OPERATING DATA REPORT DOCKET NO. 50-346 i

DATE November 9, 1983 COMPLETED BY Bilal Sarsour

-

TELEPHONE (419) 259-5000

' ext. 384 OPERATING STATUS

< 1. Unit Name: Davis-Besse #1 ,

Notes

. 2. Reportmg Period: October, 1983

3. Licensed Thermal Power (MWr):

2772

4. Nameplate Rating (Gross MWe1: 925

5. Design Electrical Rating (Net MWe): 906

6. Maximum Dependable Ca scity (Gross MWe): 918 -

l 7. Maximum De,wndable Capacity (Net MWe): 874

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

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

10. Reasons For Rest ictions.If Any:

This Month Yr to.Date Cumulative

11. Hours in Reporting Period 744 7.295 46.056

12. Number Of Hours Reactor Was Critical 698.3 5.377.7 26.273.2

13. Resctor Reserve Shutdown Hours 45.7 515.2 3,879.3

14. Hours Generstor On-Line 643.3 5,185.3 24,944.9

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gmss Thermal Energy Generated (MWH) 985.280 12.525,609 57,898,370 .

17. Gross Electrical Energy Generated (MWH) 304.074 __, 4.145.224 19,251,242

18. Net Electrical Energy Generated (MWH) 27'.989 3,903,341 18,018,781

19. Unit Service Factor 86.5 71.1 54.2

20. Unit Availability Factor 86.5 71.1 57.9

21. Unit Capacity Factor (Using MDC Net) 41.8 61.2 44.8

22. Ur.!! Capscity Factor (Using DER Net) 40.4 59.1 43.2

23. Unit Forced Outage Rate 13.5 10.0 18.6

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

i

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

26. Units in Test Status (Prior to Commercial Operation): Forecast Achie$ed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (4/77)

t DOCKET NO.

50-346 ,

UNIT SHUYDOWNS AND POW:dt REDUCTIONS UNIT NAME Davis-Besse Unit 1 DATE Ndvester 9. 1983 REPORT MONTil October, 1983 COMPLETED BY R11sai h r anier TELErilONE 419-259-5000, Ext. 384 E E

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Licensee ,E*,,

g'a., Cause & Corrective r No. Dale 5g .2 Event g,? 6?

gO Action su i 4 gE @ j g, Repor ar vi u g Prevent Recurrence 6

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) 9 83 10 02 F 24.0 A 3 N/A EB TRANSF The reactor tripped on low Reactdr Coolant System (RCS) pressure due to an attempted plant runback with con-

,} trols in manual after receiving a main transformer danger alarm. Dur-ing the startup,.the reactor tripped ,

e from approximately 1% power due to improper. operation of the #2 Startup Feedwater Valve. '

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10 83 10 03 F 23.7 A 3 N/A HH INSTRU The reactor tripped on low RCS pres-sure when the' main feedwater control valve went full open causing an overfeed condition. The erroneous

valve actuation resulted from an

Jmproperly implen.ented facility

- modification to the Integrated Con-trol System (ICS).

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F: Forced

'2 Reason:

3 Method: Exhibit G. Instruct ons I S: Schedu!cd A Equipment Failure (Explain) 1-Manual for Preparation of Data 1 B.Mainienance or Test 2 Manual Scrani. Entry Shecis for Licensee C Refueli 4g 3-Automatic Scram. Event Repos (LERIFile(NUREG-D-Regulatory Restriction 4-Continuation frorn Previous Month 0161)

I!-Operatur Training & License Examination l 5 Load Reduction

F Administrative 9-Other (Explain) $ '

G-Operational lipor (linplain) Exfiibit I Same Source l'l/77) Il Other (Explaisi;

DOCKET NO. 50-346 '. -

. UNIT SHUTDOWNS AND POhut REDUCTIO.NS Davis-Besse Unit 1

. UNIT NAME DATE ' November 9, 1983 COMPLETED BY Bilal Sarsour REPORT MONTil Orgaber , 1983 419-259-5000. Ext. 384 TELErilONE

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'r No. Date k

Eg Y

J 2g&

Y Licensee Event

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k. Cause & Corrective Action su

>- 2 fE E ;y, c'g Repost a vi U o u

Prevent Recurrence 6

l 11 83 10 08 F 16.3 A 5 N/A HA PIPEXX The turbine generator was taken off l .

line to repair a high pressure tur-bine casing drain steam leak but the reactor stayed critical.

12 83 10 15 F 22.8 A 3 NP-33-83-77 HH INSTRU During the power reduction, the

, reactor tripped by the Anticipatory Reactor Trip System (ARTS) generated j.

3 by trips in the Steam and Feedwater ji . Rupture Control System (SFRCS) due to problems with the main feed pump turbine control system.

13 83 10 29 F 13.9 A 5 N/A HA INSTRU The turbine generator was taken off~

l ; line to repair Turbine Control Valve j f3, but the reactor stayed critical, j- '

I See Operational, Summary for further

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details.

1 2 3 4 l f F: Forced 8 Reasim- Method: Exhibit G-Instructions j- S: Schedu!<d A Equipment Failure (Emplain) I-Manual for Preparation of Data

1 B Maintenance of Test 2-Manual Scrani. Entry Sheets for Licensee

! C Refueling 3-Aulomalle Scram. Event Repost (LERI File (NUREG-D Regulatory Restrictism 4-Continuation from Previous Month 016l)

F Operatur Training & License Examinalion l 5 Load Reduction F Administrative ; 9-Other (Explain) -

5 '

ll' G Operatiimal Esrur (Explain)

Il Other (Explain)

Edibit I . Sane Source

N/77)

OPERATIONAL

SUMMARY

October, 1983 10/1/83 - 10/2/83

~ Fo'llowing the completion of the refueling outage on September 30, 1983, reactor power was slowly increased and attained 40% full power on October 2, 1983.

At 0140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months on October 2, 1983, a rapid power reduction was initiated due to a main transformer trouble alarm received in the Control Koom. At 0145 hours0.00168 days 0.0403 hours 2.397487e-4 weeks 5.51725e-5 months .on October 3, 1983, and during the power reduction, a combination of existing conditions and events resulted in a mismatch of reactor power and feedwater flow that resulted in a reactor trip on low Reactor Coolant System (RCS) pressure from approximately 20% full power. The cause of the transformer difficulties was due to a defective contact in the main generator exciter field breaker interlock. This caused the transformer cooling system to fail to operate in automatic.

The reactor was critical at 1416 hours0.0164 days 0.393 hours 0.00234 weeks 5.38788e-4 months on October 2, 1983. At 1511 hours0.0175 days 0.42 hours 0.0025 weeks 5.749355e-4 months on October 2, 1983, while the unit was at approximately 1% power, a reactor trip occurred due to improper operation of the #2 Startup Feedwater Valve. The valve did not open until the steam generator level was below the low level limit, after which the valve opened excessively. This reduced the feedwater flow to Steam Generator #1, which resulted in two Steam and Feedwater Rupture Control System (SFRCS) half trips, and the Anticipatory Reactor Trip System (ARTS) automatically tripped the reactor, i

[ The reactor was critical at 1845 hours0.0214 days 0.513 hours 0.00305 weeks 7.020225e-4 months on October 2, 1983.

i 10/3/83 - 10/7/83

! The turbine generator was synchronized on line at 0143 hours0.00166 days 0.0397 hours 2.364418e-4 weeks 5.44115e-5 months on October 3, 1983.

l Reactor power was slowly increased and attained approximately 28% full

! power at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months on October 3, 1983.

At 0446 hours0.00516 days 0.124 hours 7.374339e-4 weeks 1.69703e-4 months on October 3, 1983, while the unit was at approximately 28%

full power, the main feedwater control valve went full open, causing an .

l overfeed condition which resulted in a reactor trip oa low RCS pressure.

l The problem was found to be in the Integrated Control System (ICS).

The reactor was critical at 2328 hours0.0269 days 0.647 hours 0.00385 weeks 8.85804e-4 months on October 3, 1983.

The turbine generator v. synchronized on line at 0428 hours0.00495 days 0.119 hours 7.07672e-4 weeks 1.62854e-4 months on October 4, 1983.

The reactor power was slowly increased to 40% of full power which was attained at 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months on October 4, 1983. Physics testing at the 40%

power level was completed at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months on October 7, 1983.

10/8/83 - 10/13/83 Reactor power was maintained at approximately 40% power until 0720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months on October 8, 1983, when the turbine was taken off line to r pair high pressure turbine casing drain steam leak, but the reactor stayed critical at approximately 7% power.

- The turbine generator was synchronized on line at 2337 hours0.027 days 0.649 hours 0.00386 weeks 8.892285e-4 months on October 8, 1983.

The reactor power was slowly increased and attained approximately 50%

power on October 9, 1983.

Reactor power was limited to approximately 50% power due to main feed pump turbine control problems.

Reactor power was maintained at 50% power until 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on October 13, 1983, when reactor power was increased to approximately 75% full power.

10/14/84 - 10/24/83 Reactor power was maintained at approximately 75% power until 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months on October 14, 1983, when a manual power reduction was initiated by the operator due to another high pressure turbine casing drain line steam leak.

At 0427 hours0.00494 days 0.119 hours 7.060185e-4 weeks 1.624735e-4 months on October 15, 1983, during the power reduction, a reactor trip occurred from approximately 31% of full power. Problems with the main feed pump turbine control system had caused feedwater oscillations that resulted in an ARTS trip generated by trips in the SFRCS. The ARTS trip automatically tripped the reactor.

The reactor was critical at 1530 hours0.0177 days 0.425 hours 0.00253 weeks 5.82165e-4 months on October 15, 1983. The turbine generator was' synchronized on line at 0313 hours0.00362 days 0.0869 hours 5.175265e-4 weeks 1.190965e-4 months on October 16, 1983.

! Recctor power was slowly increased and attained approximately 75% power at

! 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on October 24, 1983.

10/24/83 - 10/28/83 Reactor power was maintained at approximately 75% power until 0232 heurs l on October 26, 1983, when an automatic runback to 60% occurred because of

! the loss of Control Rod Group 3 out limit.

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Reactor power was slowly increased to approximately 75% of full power which was attained at 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on October 26, 1983. Physics testing at the 75% power level was completed at 0930 hours0.0108 days 0.258 hours 0.00154 weeks 3.53865e-4 months on October 28, 19&3.

10/29/83'- 10/31/83

Reactor power was maintained at approximately 75% of full power until 0422 hours0.00488 days 0.117 hours 6.977513e-4 weeks 1.60571e-4 months on October 29, 1983, when the turbine generator was taken off line to repair control. valve #3, but the reactor stayed critical. , ,

The turbine generator was synchronized on line at 1813 hours0.021 days 0.504 hours 0.003 weeks 6.898465e-4 months on October 29, 1983.

4 The reactor power was slowly increased and attained approximately 90%

power at 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months on October 30, 1983, and maintained at this power level for the rest of the month.

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REFUELING INFORMATION DATE: October, 1983

1. Name of facility: Davis-Besse Unit 1

2. . Scheduled date for next refueling shutdown: August 3, 1984,

3. Scheduled date for restart following refueling: October 26, 1984

4. Will refueling or resumption of operation thereafter require a .

technical specification change or other license amendment? If answer is yes, what in general will these be? .If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review ComrAttee to determine whether any unreviewed safety questions are aes.fiated with the core reload (Ref. 10 CFR Section 50.59)?

Ans: Expect the Reload Report to require standard reload fuel design Technical Specification changes (3/4.1 Reactivity Control Systems and 3/4.2 Power Distribution Limits).

5. Scheduled date(s) for submitting proposed licensing action and supporting information: July, 1984

6. Important licensing considerations associated with refueling, e.g.,

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

Ans: None identified to date.

, 7. The number of fue1' assemblies (a) in the core and (b) in the . spent fuel storage pool.

(a) 177 (b) 140 - Spent Fuel Assemblies

8. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or i is planned, in number of fuel assemblies.

Present: 735 Increase size by: 0 (zero)

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

Date: 1993 - assuming ability to unload the entire core into the spent fuel pool is maintained.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 77-107 SYSTEM: Steam and Feedwater Rupture Control System (SFRCS)

COMPONENT: Steam Generator Level Transmitters CHANGE, TEST OR EXPERIMENT: Trip setpoints for the Steam Generator level transmitters, LSL-SP9A6, SP9A7, SP9A8, SP9A9, SP9B6, SP9B7, SP9B8, and SP9B9, were changed from 24" water 5" to 23" water i 2". This change was verified February 22, 1980.

REASON FOR CHANGE: Technical Specifications require the water level to be 220" above the lower tubesheet. -The previous setpoint would have allowed the water to get as low as 19" above the lower tubesheet. The new setpoint.

allows a 1" margin from the minimum specified in the Technical Specifications.

SAFETY EVALUATION: This change has assured that the Steam Generator levels will remain within the limits set forth in the Technical Specifica-tions.

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I COMPLETED FACILITY CHANGE REQUESTS l l

1 FCR No: 77-276 'l l

) SYSTEM: Communications

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COMPONENT: N/A i

CHANGE, TEST OR EXPERIMENT:

Work implemented by this FCR was completed June 15, 1981. This involved the addition of a visual page system and a visual alarm system in the Emergency Diesel Generator rooms, day tank rooms and the Diesel Fire Pump room. The visual page system is activated by pushing and holding the button located next to the page telephone console in the Control Room.

This actuates rotar amber lights in these rooms. The visual alarm system's rotary red lights are actuated by the fire alarm, containment ,

l evacuation alarm or initiate emergency procedures alarm to warn an operator in these rooms.

REASON FOR CHANCE:

An operator could not be paged in the Emergency Diesel Generator rooms prior to this addition when the diesel was running due to the high noise lavel.

SAFETY EVALUATION:

This change has not affected the safety function of the Emergency Diesel Generator system because this modification was only to the communications systems in these rooms. This FCR was safety related only because PICAS were involved. Installation in accordance with PICA has precluded the creation of any adverse environments. This was not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-528 SYSTEM: Radiation Monitoring ,,

COMPONENT: RE2387 and RE2389 CHANGE, TEST OR EXPERIMENT: Cooling ductwork was extended from RE2004, Safety Features Actuation System (SFAS) Channel 1 Containment Radiation Monitor, to RE2387, Containment Wide Range Radiation Monitor, and from RE2007, SFAS Channel 4 Containment Radiation Monitor, to RE2389, Contain-ment Wide Range Radiation Monitor. This work was completed August 15, 1980.

REASON FOR CHANGE: The ambient air around these monitors was hotter than specified in design specifications. Cooling these monitors has reduced their failure rate which had been about once every six months.

SAFETY EVALUATION: The two 10 inch diameter stainless steel duct runs were subject to PICA requirements to assure that they have not created r.ny new adverse environments to nearby safety related equipment. No unrevie:wed safety question exists, i

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-014

- SYSTEM: Fire Protection

' COMPONENT: Auxiliary Building Fire Detection CHANGE, TEST OR EXPERIMENT:

l Fire detection systems were installed in rooms 101, 105, 115, 113, 110 and 124, all of which are on elevation 545'-00" of the Auxiliary Building.

Work was completed October 30, 1980.

REASON FOR CHANGE:

This change was completed to upgrade the Fire Protection System in order to comply with commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

Installation in accordance with the core drill report and PICA has precluded these portions from creating any new adverse environments. An unreviewed safety question was not involved.

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COMPLETED FACILITY CHANGE REQUESTS ,

FCR No: 79-020 SYSTEM: Fire Protection .,

. COMPONENT: Fire detection CHANGE, TEST'OR EXPERIMENT:

Fire detection systems were added in rooms 216, 218, 214, 215, 220, 317 and 410. All rooms'are located in conte.inment. Work was completed July

18. 1980.

REASON FOR CHANGE:

This modification was completed to upgrade the Fire Protection System in .

order to comply with commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

Installation in accordance with the core drill report and PICA has pre-cluded these portions from creating any new adverse environments. An unreviewed safety question was not. involved.

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COMPLETED FACILITY CHANGE REQUESTS ,

FCR No: 79-044 SYSTEM: Fire Protection _,

COMPONENT: Sprinkler System CHANGE, TEST OR EXPERIMENT:

A sprinkler system was added to Room 314, the No. 4 Mechanical Penetration roc:: , elevation 585'-00". Work was completed February 26, 1981.

REASON FOR CHANGE:

This change was completed to upgrade the Fire Protection System in order to comply with commitments made in the Fire Hazard Analysis Report. -

SAFETY EVALUATION:

Installation in accordance with core drill reports and PICA has precluded these portions from creating any new adverse environments. An unreviewed safety question was not involved.

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COMPLETED FACILITY CHANGE REQUESTS .

FCR NO: 79-047

. SYSTEM: Fire Protection - -

COMPONENT: Sprinkler System CHANGE, TEST OR EXPERIMENT:

A sprinkler system was added to passage 310 and hatch 313, both on elevation 585'-0". Work was completed July 23, 1980.

REASON FOR CHANGE _:

This change was required by commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

This work is non-nuclear safety related except for a core drill cut out. Installation in accordance with the "Q" core drill report-and PICA has precluded those portions from creating any new adverse environment. An unreviewed safety question was not involved.

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COMPLETED FACILITY CH/.NGE REQUESTS FCR NO: 79-048 SYSTEM: Fire Protection

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COMPONENT: Sprinkler System CHANGE, TEST OR EXPERIMENT:

A sprinkler system was added to Room 304, the east-west corridor on elevation 585'-00". Work was completed November 11, 1980.

REASON FOR CHANGE:

This change was completed to upgrade the Fire Protection System in order to comply with commitments made in the Fire Hazard Analysis Report. ,

SAFETY EVALUATION:

This work is non-nuclear safety related except for 2 core drill cutouts.

Installation in accordance with the "Q" core drill report and PICA has precluded these portions from creating any new adverse environments. An unreviewed safety question was not involved.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-049 SYSTEM: - Fire Protection COMPONENT: Sprinkler System CHANCE. TEST OR EXPERIMENT:

A sprinkler system was added for the radwaste' exhaust equipment and-

! Main Steam exhaust fan room, room 561 on elevation 623'-00". Work was completed September 28, 1980.

REASON FOR CHANGE:

This change was completed to upgrade the Fire Protection System in order ,

to comple with commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

This work is non-nuclear safety related except for 2 core drill cutouts.

Installation in accordance with the core drill report and PICA has' precluded these portions from creating any new adverse environments. An unreviewed safety question is not involved.

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COMPLETED FACILITY CHANGE REQUESTS P FCR NO: 79-053

. SYSTEM: Fire Protection

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COMPONENT: Sprinkler System CHANGE TEST OR EXPERIMENT:

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A sprinkler system was added to the clean waste receiver tank area, Room 124 on elevation 565'-00". Work was completed June 5, 1980.

REASON FOR CHANGE:

This change was completed to upgrade the Fire Protection System in order to comply with commitments made in the Fire Hazard Analysis Report. ,

SAFETY EVALUATION:

Installation in accordance with core drill report and PICA has precluded

' these portions from creating any new adverse environments. An unreviewed safety question was not involved.

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COMPLETED FACILITY CHANGE REQUESTS .

FCR No: 79-063

. SYSTEM: Fire Protection --

COMPONENT: Hose Station CHANGE, TEST OR EXPERIMENT:

A hose station was added in the Diesel Generator Room 319, on elevation 585'-00". Work was completed February 26, 1981.

REASON FOR CHANGE:

This change was comp.teted to upgrade the Fire Protection System in order to comply with commitments in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

This work is non-nuclear safety related except for the installation of the core drill. Installation in accordance with the core drill procedure has precluded the creation of any adverse environments. This was not j an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS .

FCR NO: 79-238 SYSTEM: Fire Protection ,,

COMPONENT: Sprinkler System CHANGE, TEST.OR EXPERIMENT:

This FCR was implemented to add a sprinkler system in storage room 405 at elevation 603'. All work was completed August 25, 1980.

REASON FOR CHANGE:

This change was required by commitments made in the Fire Hazard .

Analysis Report.

SAFETY EVALi?ATION:

Work required by this FCR is non-nuclear safety related except for 1 core drill cut.out through a negative pressure boundary. Installation in accordance with the "Q" core drill report and PICA requirements pre-cluded the creation of any adverse environments.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-239

. SYSTEM: Fire Protection COMPONENT: Sprinkler System CHANGE, TEST OR EXPERIMENT:

A sprinkler system was added to room 427, No. 2 Electrical Penetration Room on elevation 603'-0". Work was completed October 8, 1980.

REASON FOR CHANGE:

? This change was completed to upgrade the Fire Protection System in order ,

to comply with commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

Installation in accordance with core drill report has precluded these portions from creating any new adverse environments. An unreviewed safety question was not involved.

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-COMPLETED FACILITY CHANGE REQUEST FCR NO:- 79-240 SYSTEM: Fire Protection I r

, COMPONENT: Sprinkler System -- ,

CHANGE, TEST, OR EXPERIMENT:

A sprinkler system was added in No. 1 Electrical Penetration Room 402, Eleva-

' tion 603'-0". Work was completed September 28, 1980.

. REASON FOR CHANGE:

This modification was completed to upgrade the fire protection system in order to comply with commitments made in the Fire Hazard Analysis Report.

SAFETY EVALUATION:

This work was non-nuclear safety related except for a core drill. Installa-tion in accordance with the core drill report and PICA has precluded those portions from creating any new adverse environments. An unreviewed safety question was not involved.

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COMPLETED FACILITY CHANGE REQUESTS .

FCR NO: 79-325

. SYSTEM: Communications' -- -

COMPONENT: N/A CHANGE, TEST OR EXPERIMENT:

New speakers for the "Gaitronics" were added in the Boric Acid Evaporator-rooms, 234 and 235. Work was completed April 28, 1980.

REASON FOR CHANGE:

~

NRC Bulletin 79-18 requires the communications and alarm system to be '

clearly audible throughout the entire plant. The alarms could not be heard in these rooms since they had no speakers and have a high backround noise level..

SAFETY EVALUATION:

- The conduit for this addition was routed through an existing penetration which was resealed with silicone foam. The implementation of this FCR has had no effect on plant safety. This was not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS -

FCR NO: 82-093

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SYSTEM: Reactor Coolant System (RCS)

COMPONENT: Pressurizer Surge Line CHANGE, TIST OR EXPERIMENT:

This change involved the modification of the gaps between the pressurizer surge line and pipe whip restraints SL3, SL4, SL6 and SL7.

All work was completed and Bechtel drawing 7749-C-189 was updated July 31, 1983.

REASON FOR CHANGE:

Toledo Edison Non-conformance Report 397-82 noted that certain gaps between the praasurizer surge line and the slims on the pipe whip res-traints SL1 thruagh SL8 were different than shown on drawing 7749-C-189.

The as measured gaps were analyzed and it. was determined that four gaps needed to be changed and that the others were acceptable but the design drawing needed to be changed.

SAFETY EVALUATION:

The function of pipe whip restraints is, in the event of a pipe rupture, to prevent the ruptured pipe from whipping around and damaging other safety related piping and equipment. The energy that a pipe whip restraint is designed to withstand is a function the gap between the pipe and the restraint. Gaps that were larger than shown on Bechtel drawing 7749-C-189 were modified to conform to the design drawing gaps that were smaller than shown on the drawing are adequate for pipe whip design.

These smaller gaps were also renewed for seismic and thermal movement and found to be acceptable. This modification has not created an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST ,

FCR NO: 82-127 SYSTEM: 480 Volt Essential AC Power _,

COMPONENT: Heater' Breakers BE1223 and BF1217 CHANGE, TEST OR EXPERIMENT: This FCR implemented the replacement of existing 200 amp trip units on essential heater breakers BE1223 and BF1217 with 250 amp trip units. These breakers were installed and trip points were verified on August 31, 1982.

REASON FOR CHANGE: The current through the essential heater breakers BE1223 and BF1217 is approximately 160 amps, which is normal for a full .

heater bank. With a trip setpoint of 200 amps, the normal current, rather than an overcurrent, was causing these units to trip. The non-essential heater breakers were replaced with 250 amp trip units by FCR 78-430.

SAFETY EVALUATION 1 The safety function of the Class IE breakers is to timely isolate the non-1E pressurizer heaters from the IE power supply in case of electrical fault downstream from the breaker. Replacement of these trip units enhances this function rather than adversely affects it.

Therefore, this is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS -

FCR NO: 83-016

. SYSTEM: Main Steam -' '

COMPONENT: MS611, MS603 CHANGE, TEST OR EXPERIMENT:

This FCR was implemented to change the torque switch settings for motor operated valves MS611, Steam Generator (SG) 1-1 drain line to low pressure condenser motor isolation valve and MS603 SG 1-2 drain line to high pressure condenser motor isolation valve. The new settings for both MS611 and MS603 are 3.0 to open and 1.5 to close. Work was completed January 29, 1983.

REASON FOR CHANGE:

Torrey Pines Technology had recommended these torque switch settings in their limitorque motor operated valve study to reduce the likelihood of failure of these valves.

SAFETY EVALUATION:

The safety function of these valves is to provide assurance that the con-tainment atmosphere is isolated from the outside atmosphere in the event of a release of radioactive material to the containment atmosphere or pressurization of containment. The modification has enhanced the relia-bility of these valves. The safety function of valves MS611 and MS603 were not adversely affected, therefore, no unreviewed safety question existed. .

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m TOLEDO

%ms EDISON November 9, 1983 Log No. KS3-1583 File: RR 2 (P-6-83-10)

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- Docket No. 50-346 L,icense No. NPF-3 Mr. Norman Haller, Director Office of Management and Program Analysis U. S. Nuclear Regulatory Commission 7ashington, D.C. 20555

Dear Mr. Haller:

6 Monthly Operating Report, October 1983 Davis-Besse Nuclear Power Station Unit 1 Enerlosed are ten copies of the Monthly Operating Report for Davis-Besse Nuc1 car Power Station Unit 1 for the month of October, 1983.

If you have any questions, please feel free to contact Bilal Sarsour at (419) 259-5000, Extension 384.

Yours truly, n w y o. q / a q Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station -[7[2 I TDM/BMS/ljk I)\

Enclosures cc: Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. Richard DeYoung, Director, w/2 Office of Inspection and Enforcement Mr. Walt Rogers, w/1 NRC Resident Inspector e

THE TOLEDO EDISON COMPANY EOISON PLAZA 300 MAOISON AVENUE TOLEDO. OHIO 43652

_ , _ _ . , , . - _ . . _, _ _ . _ . - .

ML20082S956

Text

SUMMARY

Dear Mr. Haller:

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