ML20024F041

Monthly Operating Rept for Jul 1983.W/830808 Ltr

ML20024F041
Person / Time
Site:	Davis Besse
Issue date:	08/08/1983
From:	Murray T, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC
References
K83-1101, NUDOCS 8309080145
Download: ML20024F041 (21)

Text

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AVERAGE DAILY UN T POW 5R LEVEL DOCKET NO.

UNIT Davis-Besse Unit 1 DATE luzust 8, 1983 1

COMP!,ETED BY Bilal Sarsour I (419) 259-5000, TELEPHONE i

Ext, 384 July, 1983 ,

MONTH ,

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL '

(MWe Net) '(MWe-Net)

I 781 617-37 2 707 gg 606 3 650 19 598 4 . 635 20 596 3 627 21 602 6 629 592 22 7 627 23 600 g 630 24 598 9 629 25 289 10 630 26 0 11 628 27 0-12 622 2g 0 13 623 29 0 14 620 0 30 15 619 0 3:

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l INSTRUCTIONS On this forrnat. list the average daily unit power levelin MWe Net for each day in the reporting snunth.Cornpute to the nearest whole megawatt.

(9/77)

B309080145 830808 FDR ADOCK 05000346 R PDR 6%q

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OPERATING DATA REPORT DOCKET NO. 50-346 .

- DATE Anoust 8. 1983 COMPLETED BY Bilal sarsour TELEPHONE 419-259-5000,

, Ext. 384 OPERATING STATUS _

Davis-Besse Unit 1 Notes

l. Unit Name:

. 2. Reporting Period: July, 1983

3. Licensed Thermal Power (MWr): 2772 ,

4. Nameplate Rating (Gross MWe): 925

5. Desagn Electrical Rating (Net MWe): onA

6. Maximum Dependable Capecity (Gross MWe): 918 -

7. Maximum Dependable Capacity (Net MWe)

874

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):

10. Reasons For Restrictions.lf Any:

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This Month Yr.-to-Date Cumulative

II. Hours In Reporting Period 744.0 5,087.0 43,848.0

12. Number Of Hours Reactor Was Critical 588.4 4,591.2 25,486.7

13. Reactor Reserve Shutdown Hours 155.6 469.5 3,833.6

14. Hours Generator On-Line 588.4 4,538.8 24,298.4 IS. Unit Reserve Shutdown Hours 0.0 0.0 1.732.5

16. Gross Thermal Energy Generated (MWH) 1.202.826 11.534.142 56,906,903 .

17. Gross Electrical Energy Generated (MWH) 394.838 ,_, 3.841.150 18,946,804

18. Net Electrical Energy Generated (MWH) 366.862 3,631,352 17,746,792 l 19. Unit Service Factor 79.1 89.2 55.4

20. Unit Availability Factor 79.1 89.2 59.4

21. Unit Capacity Factor (Using MDC Net) 56.4 81.7 46.3

22. Unit Capacity Factor (Using DER Net) 54.4 78.8 44.7

23. Unit Forced Outage Rate 20.9 10.8 18.9

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

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- 2S. If Shut Down At End Of Report Period. Estimated Date of Startup: September 20, 1983

26. Units in Test Status (Prior to Commercial Operation): Forecast Achiesed INITIAL CRITICALITY INITIA L ELECTRILITY COMMERCIA L OPER ATION I N/77)

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DOCKET NO.

50-346 ',

- UNIT SHUTDOWNS AND POWNI REDUCTIONS Davis-Besse Unit l' -

UNIT NAME DATE August 8, 1983 COMPLETED BY Bilal Sarsour REPORT MONTil July 1983 TELErilONE (419) 259-5000 Ext. 384 t .

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ge, cause & Corrective .

t No. Date k n Ej .s j s c2 Event gg 93 Action to y

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7- 83 07 25 F 155.6 A 3

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The reactor tripped due to Steam and Feedwater Rupture Control System

, (SFRCS) actuation.

. A unit outage was initiated to per-

. form scheduled maintenance and

• refueling work.

4 l . See Operational Summary for further details.

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l i 1 2 3 4 l F: Forced Reason: Method: Exhibit G Instructions j j S: Schedu!cd .4-Equipn.ent Failure (Explain) l-Manual for Preparation of Data j 3-Maintenance of Test 2-Manual Scram,. Entry Sheets for Licensee i C Refueling 3-Automatic Scram. Event Report (LERI File (NUREG-

9. Regulatory Restriction 4-Continuation from Previous Month - 0161)

E-Operator Training & License Examination 5-Load Reduction F Administrative l 9 Other (Explain)

G Operational Es ror (Explain) Extiibit I Same Source 08/77) Il-Other (Explain)

OPERATIONAL

SUMMARY

JULY, 1983 7/1/83 - 7/24/83: Reactor power was maintained at approximately 90 percent power until 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on July 2, 1983, when power was slowly reduced and attained approximately 73 percent power on July 5, 1983. The power reduction was due to low load requirements over the holiday.

6 Reactor power was maintained at 73 percent power until 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> on July 17, 1983 when a manual reduction of reactor power was initiated due to the end of cycle coastdown.

7/25/83 - 7/31/83: On July 25, 1983, while the unit was at approximately 70 percent power, a reactor trip occurred due to Steam and Feedwater Rupture Control System (SFRCS) actuation.

The reactor tripped by the Anticipatory Reactor Trip System (ARTS) during the replacement of the power supply for SFRCS Channel 3. The power supply failed at 0319 hours0.00369 days <br />0.0886 hours <br />5.274471e-4 weeks <br />1.213795e-4 months <br /> on July 25, 1983, causing a half trip of SFRCS Channel 1. At 1226 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.66493e-4 months <br /> on July 25, 1983, Channel 1 produced a full trip that tripped the reactor.

A unit outage was initiated to perform scheduled maintenance and refueling work. Full details on the work items performed during the scheduled maintenance outage will be presented in next month's operational summary.

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REFUELING INFORMATION DATE: July, 1983

1. Name of facility: Davis-Besse Unit 1

2. Scheduled date for next refueling shutdown: July 25, 1983

3. Scheduled date for restart following refueling: September 20, 1983

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 Committee to determine whether any unreviewed safety questions are associated 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, 1983

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 fuel assemblies (a) in the core and (b) in the spent fuel storage pool.

(a) 177 (b) 92 - 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 is planned, in number of fuel assemblies.

Present: 735 Increase size by: 0 (zero)

9. The projected date 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.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-029 SYSTEM: . Auxiliary Building Emergency Lighting, Elevation 643 COMPONENT: N/A CHANGE, TEST, OR EXPERIMENT: This FCR called for the installation of emergency lighting in rooms 600 (Purge Inlet Equipment Room), 601 (#1 Main Steam Line Area),

and 602 (#2 Main Steam Line Area). Work was completed January 3, 1980.

REASON FOR CHANGE: This change was initiated to comply with commitments made

. in the Fire Hazard Analysis Report.

SAFETY EVALUATION: This change was nuclear safety related because it called for

. the installation of emergency lighting supports on nuclear safety related walls.

Installation in accordance with PICA requirements ensured no new adverse environ-ments were created.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-168 SYSTEM: Borated Water Storage Tank (BWST) i COMPONENT: High pressure injection recirculation line heat trace circuits 17, 20, 21, and 24 CHANGE, TEST, OR EXPERIMENT: Work implemented by FCR 79-168 was ccupleted March 27, 1981. Redundant heat trace circuits 17 and 21 were shertened to provide freeze protection only for the portion of the BWST minimum circula-

. tion pipe within the pipe tunnel. The thermocouples for these circuits were then relocated as appropriate.- Redundant heat trace circuits 20 and 24 were then added to protect the portion of the pipe outside of the pipe' tunnel from freezing. The thermocouples for these circuits were located near the top of the BWST.

REASON FOR CHANGE: The portion of pipe HCC-91 above ground level froze Janu-ary 3, 1979. This rendered High Pressure Injection Pump 1-1, which was being tested, inoperable due to a low recirculation flow indication. Reference Licensee Event Report NP-32-79-03 (79-034).

SAFETY EVALUATION: This change has not adversely affected the function of the high pressure injection pump or BWST minimum circulation pipe HCC-91. It has enhanced the operation of the minimum circulation line by adding more reliability to ensure that the line does not freeze during cold weather. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-199 SYSTEM: Control Room Heating, Ventilating, and Air Conditioning COMPONENT: Damper Shafts HV5301 A and B and HV5311 A and B CHANGE, TEST OR EXPERIMENT: This FCR, which was completed May 2, 1980, involved the replacement of the shafts in the Control ifoom isolation damper. The original shafts were carbon steel and were replaced by stainless steel.

REASON FOR CHANGE: This change was recommended by the damper manufacturer because moisture, which was accumulating on the shaf t at the bushings, was causing the dampers to bind in one position. This moisture was also corroding the shaft at the shaft / bushing contact points.

SAFETY EVALUATION: The Control Room dampers are part of the Control Room Emergency Ventilation System which must remain operable in Modes 1 thrcugh

4. This modification has enhanced the operation of the dampers since the stainless steel is much more durable and will prevent corrosion.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-360 SYSTEM: Reactor Coolant System COMPONENT: Pressurizer -

CHANGE, TEST OR EXPERIMENT: This FCR provided for the seal welding of the

pressurizer heater bundle diaphragas to the pressurizer' heater belt. This seal weld is not a pressure boundary. Work was completed May 21, 1981.

REASON FOR CHANGE: This was recommended by Babcock and Wilcox due to problems of leakage in this area.

SAFETY EVALUATION: This seal has lessened the likelihood of leakage from this area. This change not has affected the operation of the pressurizer and, therefore, does not constitute an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-190 SYSTEM: N/A COMPONENT: Snubber Hangers PS-H7 and 30-GCC-8-H5 CHANGE, TEST, OR EXPERIMENT: This FCE involved the modification of two hangers to permit the lengthening of the cold piston settings of the snubbers on these hangers. Work was completed September 7, 1980.

REASON FOR CHANGE: The previous cold piston settings of these snubbers were below the minimum cold piston setting allowed by Bechtel drawing 12501-M-618.

The allowable cold piston setting range for PS-H7 is 1-5/16" to 4-1/8". The actual cold piston setting of this snubber was found to be 15/16" or 3/8" below minimum. The cold piston setting range for 30-GCC-8-H5 is 1-3/8" to 5-1/8".

Snubber 30-GCC-8-H5 had a cold piston setting of 1-7/32" or 5/32" below minimum.

SAFETY EVALUATION: The snubber setting in hangers PS-H7 and 30-GCC-8-H5 were within the 1/2" tolerance but the snubbers did not bottom out. The as-found condition was evaluated by Bechtel and determined acceptable for short term operation, reference BT-13402. Operability of the snubbers has been enhanced by this change. No unreviewed safety question exists.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-396 SYSTEM: Reactor Coolant System C0tTONENT: Pressurizer Liquid Sampling Valve RC239B Torque Switch CHANGE, TEST OR EXPERIMENT: FCR 79-398 was issued to revise electrical drawing EIS to show the new values of the torque switch' settings for RC239B. The new torque switch settings are 1.0 to close and 2.0 to open.

This change was completed January 12, 1983.

REASON FOR CHAJG1: Previous maintenance had determined that the torque values were tosufficient to operate the valve.

SAFETY EVALUATION: The safety functions of the pressurizer liquid sample valve are:

1) To provide a path for sampling of the Reactor Coolant System

2) To provide double isolation between the two types of piping classes The safety functions of valve RC239B were not adversely affected and, therefore, no unreviewed safety question was involved.

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COMPLETFSFACILITY CHANGE RFOUEST

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SYSTEM: Doors' . '

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COMEONENT: fWire voverr* barricades ,

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CHANGE, TEST OR EXPERIMENT: Undcr,this FCR the following

• permanent wir.e woven barricades and doorways were, ins'talled: s - -

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Door No. - Elev. gpeription } '

136 537' -0" Access Tunnel 1 230 565' -0" Access Tunnel s 231 567' -6" Access Tunnel 232 568' -0"' Flooding Tank Area 233 565' -0" Flooding Tank Area N/A 579'~-0" Over Transfer Tubes 370 585' -0" To Incore Monitor Tank N/A . '545' -0" Normal Sump to Under Vessel 372 ,365' -0" Annulus N/A 585',-0" To Transfer Tubes N/A 585'_., To Transfer Tubes The installation was ccmpleted April 17,.4981.

REASON FOR CHANGE: This will provide permanent wire woven barricades and doorways to restrict access to high radiation areas caused when the incores are out of the reactor while fuel is being transferred through

. transfer tubes.

SAFETY EVALUATION: No new adverse environments have been created by this i

installation. No unreviewed safety question exists.

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% COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-021'

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SYSTEM: N/A COMPONENT: sWalls 4837, 4847, 4857, 4806, 4817, and 4826 CHANGE, TEST OR EXPERIMENT: This FCR removed masonry walls 4837, 4847,

. 4857, 4806, 4817, and 4826 which previously surrounhed two structural

"' columns in Electrical Penetration Room 2 (Room 427, elevation 603'). The

, sceel columns were then fireproofed and existing safety related conduits were relocated to satisfactory supports. Work was completed June 3, 1982.

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REASON FOR CHANGE: Block wall reanalysis, required by Nuclear Regulatory

- - , Commission IE Bulletin 80-11, had shown that during a seismic event the walls mentioned above could develop stresses that would result in wall failure. Reference Licensee Even't Report NP-32-80-17 (80-091).

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SAFETY EVALUATION: These walls functioned as fire barriers and supports

for the safety related conduits which either penetrated or were attached s to the walls. Failure of these walls would result in loss of fire protection for the steel columns and potential damage to the attached safety related conduits.

This modification has not reduced the fire rating of the columns and has J not affected the safety related conduits in any adverse manner.

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COMPLETED FACILITY CHANGE REQUEST

_FCR NO: 81-022 '

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COMPONENT: Wall 4647 ,

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,m CHANGE TEST '0$ EXPERIMENT: This FCR added a one inch seismic joint and

' joint filler material.at tha. south end of masonry wall 4647 which separates theicable spreading room from the upper part of corridor 404 on elevation 616' of the Auxiliary Building. Work was completed Octobet 5, 1981. '

Ol '<- ) ,' REASON FOR CHANGE: Block wall reanalysis, required by Nuclear Regulatory

< 1 m. Coussission IE Bulletin 80-11, had shown that during a seismic event, localized I '.,. 4 / crushing of the masonry could occur due to the absence of the seismic joint.

/~ .'-/ This change is~.~also corrective action for Deviation Report 81-079.

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- ' SAFETY EVALUAT10': N Wall 4647 functions as a fire wall, part of the negative pressure boundary, and a support for safety related conduits which penetrate '

this wall..T This modification lowered the stresses in the masonry to the allowable limits specified in Section 3.8 of the Final Safety Analysis Report.

i - The modification .was reviewed and found not to affect the wall's ability to function,as a fire wall.. negative pressure boundary, or as a support for the

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-025 SYSTEM: N/A COMPONENT: Wall 3407 -

CHANGE TEST, OR EXPERIMENT: Four steel braces were added to the web of the floor beam attached to the top of wall 3407 to stiffen it. This wall separates Compon-ent Cooling Water Heat Exchanger and Pump Room #328 from stairway AB-1 at eleva-tion 585'. This modification was completed October 16, 1981.

REASON FOR CHANGE: Nuclear Regulatory Commission Bulletin 80-11 required block wall re-analysis which showed that during a seismic event the floor beam above wall 3407 could become overstressed. This FCR is also corrective action for Devia-tion Report 81-109.

SAFETY EVALUATION: This change has reduced the possible stresses in the floor beam above wall 3407 due to seismic loading to within the limits specified in Section 3.8 of the Final Safety Analysis Report.

This modification has not adversely affected the stability of the floor beams above wall 3407 or the wall's ability to function as a fire barrier.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-027 SYSTEjM N/A COMPONENT: Wall 5207 .

CHANCE. TEST. OR EXPERIMENT: Work implemented by this FCR was completed Septem-ber 28, 1982. This involved the attachment of steel struts to the top of wall 5207 at four locations. Wall 5207 separates Computer Room 510 from Control Cabinet Room 502 at elevation 623'.

REASON FOR CHANGE: Block wall re-analysis, required by Nuclear Regulatory Commission IE Bulletin 80-11,had shown that during a seismic event, this wall could overstress the floor beam above this wall. This is also corrective action for Deviation Report 81-111.

SAFETY EVALUATION: This modification has lowered the stresses in the floor beam above wall 5207 due to seismic loading to within the allowable limits established in the Final Safety Analysis Report for Class 1 structures. The floor system above wall 5207 and this wall's ability to function as a fire barrier have not been adversely affected by this change. This 10 not an unroviewed safety question.

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P COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-071 SYSTEM: 125 VAC Essential Power s

COMPONENT: Station Battery Chargers CHANGE, TEST, OR EXPERIMENT: The station battery charger output filter fuses were changed from 25 amps AFS25 to 60 amps AFS60. This was completed Augt.st 12, 1981.

REASON FOR CHANGE: This change will ensure maximum output filtering while main-taining the noise levels on the 125 VAC system at the lowest level. This modifi-3 cation was reconsnended by Cyberex, Inc.

SAFETY EVALUATION: The safety function of the fuse is to isolate a faulty filter capacitor bank from the DC system without total loss of output from the battery charger. All modifications were internal to the 125 VAC system and will not pre-vent the safe shutdown of the plant or create any new adverse environments.

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r COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-146 SYSTEM: 240 Volt Motor Control Centers COMPONENTi Transformers YE2 and YF2 CHANGE, TEST OR EXPERIMENT: Work implemented by FCR 81-146 was completed September 27, 1982. This involved the replacement of e'xisting 480/240 volt motor control center 3KVA transformers YE2 and YF2, previously located inside motor control centers YE2 and YF2, respectively, with 30 KVA transformers which were mounted outside of the motor control centers.

REASON FOR CHANGE: Increasing the capacity of these transformers has decreased the voltage drop due to the motor startup current. This modi-fication will provide additional assurance that the containment air sampling isolation valves close within the time requirements of Technical Specification 3.6.3.1. ---

SAFETY EVALUATION: The safety function of these transformers is to supply power to 240 and 120 volt motors actuated by safety signals.

This change insures availability of more than minimum starting voltage on motor control centers YE2 and YF2 to feed the Class 1E motors during a simultaneous start of safety actuation loads on all Class 1E buses. This is the desired result of this change and, therefore, will not create an adverse environment. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-310 SYSTEM: 13.8 KV and 4.16 KV ,

COMPONENT: Switchgear CHANCE. TEST OR EXPERIMENT: FCR 81-310 implemented the. adjustment of the contact gap on all ITH type instantaneous ground relays on the 4.16 KV and 13.8 KV feeders. The gap adjustment was increased to 0.035 inches with tolerances of +0 and -0.005 inches. Work was completed June 9, 1982.

REASON FOR CHANGE: Prior to this modification, construction personnel had caused a mechanical shock to sensitive ITH relays causing a loss of essential buses E2, YAR, and YAU; reference Licensee Event Report NP-33-81-44 (81-037). Increasing the contact gap on this type of relay has decreased their sensitivity to mechanical shock.

SAFETY EVALUATION: This safety function of the ITH relays is to trip the feeder breaker in case of a ground fault. This change has increased the reliability of the 4.16 KV and 13.8 KV distribution systems. It has not affected relay pickup.

Safety functions were not affected, therefore, there is no unreviewed safety question. ,

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• COMPLETED FACILITY CHANGE REQUEST FCR No: 82-112 SYSTEM: Auxiliary Feedwater System COMPONENT: Motor Operated Valves MV38690, MV38700, MV38710, .and MV38720 CHANGE, TEST, OR EXPERIMENT: FCR 82-112 was completed July 29, 1982. This involved increasing the open torque dial settings for the above valves from 2.75 to 3.00. The closed torque dial settings were decreased from 2.75 to

. 1.50.

REASON FOR CHANGE: The previous torque dial settings had caused a motor failure during operation. This change was specified in the Torrey Pine Tech-nology report on limitorque motor operated valves to reduce the likelihood of motor failure.

SAFETY EVALUATION: The safety functions of these valves are:

MV38960 - Provides auxiliary feedwater to Steam Generator #2 in the event of a steam line break on Steam Generator #1.

MV38710 - Provides auxiliary feedwater to Steam Generator #1 in the event of steam line break on Steam Cenerator #2.

MV38700 - Provides auxiliary feedwater to Steam Generator #1 in case of low steam generator ?evel in any steam generator, or a loss of four Reactor Coolant Eumps.

MV38720 - Provides auxiliary feedwater to Steam Generator #2 in the event of low steam generator level in any steam generator or a loss of four Reactor Coolant Pumps.

This change has increased the reliability of operation while not affecting the safety functions of these valves.

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O TOLEDO

% sus EDISON August 8, 1983 Log No. K83-1101 File: RR 2 (P-6-83-07)

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

Dear Mr. Haller:

Monthly Operating Report, July, 1983 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of July, 1983.

Yours truly, Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/BMS/1jk Enclosures cc: Mr. James C. Keppler '

Regional Administrator, Region III Enc 1: 1 copy Mr. Richard DeYoung, Director Office of Inspection and Enforcement Encl: 2 copies Mr. Walt Rogers NRC Resident Inspector Enc 1: 1 copy

,rd THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652 I\

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