ML20132F776

Monthly Operating Rept for Jul 1985

ML20132F776
Person / Time
Site:	Davis Besse
Issue date:	07/31/1985
From:	Khazarai M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K85-1193, NUDOCS 8510010410
Download: ML20132F776 (20)

Text

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b AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit 1 DATE August 9, 1985 COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000 Ext. 290 MONTH July 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) 1MWe-Net) 0 1

g7 0

2 0

gg 0

3 0

39 0

4 0

20 0

5 0

0 21 6

0 0

22 7

0 23 0

8 0

24 0

9 0

0 25 10 0

26 0

1I O

27 0

12 0

0 28 13 0

0 29 14 0

30 0

15 0

0 3,

16 0

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

(9/77) 8510010410 850731 PDR ADOCK 05000346 R

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE Aucust 9, 1985 COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

2. Reporting Period:

July 1985

3. Licensed Thermal Power (MWt):

2772

4. Nameplate Rating (Gross MWe):

915

5. Design Electrical Rating (Net MWe):

906

6. Maximum Dependable Capacity (Gross MWe):

904

7. Maximum Dependable Capacity (Net MWe):

860

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

To obtain an acceptable offset from Reactor Protection System setpoint limit

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

10. Reasons For Restrictions. If Any:

This Month Yr.-to-Date Cumulative II. Hours In Reporting Period 744.0 5,087.0 61,392.0

12. Number Of Hours Reactor Was Critical 0.0 2,845.6 35,877.I

13. Reactor Reserve Shutdown Hours 0.0 44.7 4.058.8 14.~ Hours Generator On Line 0.0 2,730.5 34,371.8

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 0.0 6,312,178 81,297,600

17. Gross Electrical Energy G nerated (MWH) 0,. 0 2,087,278 26,933,622

18. Net Electrical Energy Generated (MWH) 0.0 1.942,921 25,233,177

19. Unit Service Factor 0.0 53.7-56.0-

20. Unit Availability Factor 0.0 53.7 58.8

21. Unit Capacity Factor (Using MDC Net) 0.0 44.4 47.8

22. Unit Capacity Factor (Using DER Net) 0.0 42.2 45.4

23. Unit Forced Outage Rate 100.0 33.2 18.9

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

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

Unknown i

26. Units in Test Status (Prior Io Commercial Operation):

Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPER ATION i

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(4/77 )

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DOCKET NO.

50-346 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit 1 DATE August 9, 1985 COMPLETED BY' Morteza Khazrai REPORT MONTH July'1985 TELEPHONE (419) 249-5000, Ext. 290 "u

E "c

Licensee s,s g

g Cause & Corrective Event Action to g

gg g

guy gg g, g ug Report #

gu eo Prevent Recurrence No.

Date s

gg y

Mmf av u

o 7

85 06 09 F

744 A

4 LER 85-013 JK SC The unit remained shutdown following the reactor trip on June 9, 1985.

See Operational Summary for further details.

I I F: Forced Reason:

Method:

Exhibit G - Instructions S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling

'3-Automatic Scram Event Report (LER) File D-Regulatory Restriction 4-Continuation from (NUREG-0161)

E-Operator Training & License Examination Previous Month F-Administrative 5-Load Reduction G-Operational Error (Explain) 9-Other (Explain)

Exhibit I - Same Source (9/77)

H-Other (Explain)

OPERATIONAL

SUMMARY

JULY 1985 The unit remained shutdown the entire month of July following the reactor trip on June 9, 1985.

Investigations of the causes of the event and corrective actions continue. See NUREC 1154 for further details.

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REFUELING INFORMATION DATE: July 1985 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown:

Spring, 1986 3.

Scheduled date for restart following refueling:

Summer, 1986 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: Winter, 1985 l

6.

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

i new or different fuel design or supplier, unreviewed design or l

performance analysis methods, significant changes in fuci design, new l

operating procedures.

Ans: None identified to date.

7.

The number of fuel assemblics (a) in the core and (b) in the spent fuel storage pool.

(a) 177 (b) 204 - 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:

1992 - assuming ability to unload the entire core into the spent fuel pool is maintained.

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l COMPLETED FACILITY CHANGE REQUEST FCR No:

78-477 SYSTEM: Borated Water Storage Tank COMPONENT: LT-1525A and LT-1525C CHANGE. TEST OR EXPERIMENT:

FCR 78-477 installed a permanent shelter to protect Borated Water Storage Tank (BWST) level transmitters LT-1525A and LT-1525C and their associated source lines. This shelter was designed per Seismic Category I criteria. Work was completed February 15, 1985.

REASON FOR CHANGE: This structure was installed to prevent the BWST level transmitters, LT-1525A and LT-15250, from freezing during the winter months. The original heat tracing on the associated lines to the transmitters and in the transmitter housings was inadequate for freeze protection, requiring temporary sheltering to prevent freezing.

SAFETY EVALUATION

SUMMARY

The safety function of the BWST is to provide a water source for the Emergency Core Cooling System and the Containment Spray System during a Loss of Coolant Accident. Level transmitters LT-1525A and LT-1525C are used to measure the water level in the BWST.

The above change was made to allow the intended safety function of the BWST to be satisfied. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-178 SYSTEM: Reactor Coolant System COMPONENT: Reactor Coolant Pumps CHANGE, TEST OR EXPERIMENT: This FCR performed a 10CFR50.59 review and changed the interlock setting for starting the fourth Reactor Coolant Pump (RCP). Originally, the interlock setting required that reactor power be less than 22% before restarting the fourth RCP. This FCR review resulted in the interlock being changed to a value that would allow the fourth RCP to be started provided that reactor power was less than 60% full power.

Work was completed August 13, 1979.

REASON FOR CHANCE: This change was the result of a concern that starting the fourth RCP would cause a pressure spike in the Reactor Coolant System (RCS) that would result in a high RCS pressure trip of the Reactor Protec-tion System and the reactor. The analysis done by Babcock and Wilcox for the ef fects of starting the pump at different power levels for both the beginning of life and end of life core conditions led them to recommend the interlock settings be changed to 60% full power.

SAFETY EVALUATION

SUMMARY

This FCR does not constitute an unreviewed safety question because:

1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, previously evaluated in the FSAR, has not been increased.

2) The possibility of an accident or malfunction of a different type

.other than any evaluated previously in the FSAR has not been created.

3) The margin of safety as defined in the basis for any Technical Specification has not been reduced.

l COMPLETED FACILITY CHANGE REQUEST l

FCR No: 79-221 SYSTEM: Steam and Feedwater Rupture Control System i

COMPONENT: Startup Feedwater Pump CHANGE, TEST OR EXPERIMENT: This FCR made two changes to the Steam and Feedwater Rupture Control System (SFRCS).

First, a logic change was made to the SFRCS to allow for the feeding of the steam generators with the Startup Feedwater Pump (SUFP). Second, this FCR provided a means to manually open the following valves if actuation of SFRCS has closed these valves.

Valve No.

Description SP7A Main Feedwater 2 Startup Control Valve SP7B Main Feedwater 1 Startup Control Valve FW 601 Main Feedwater 2 Stop Valve FW 612 Main Feedwater 1 Stop Valve Originally, the actuation of SFRCS closed the valves listed above. This FCR allows the valves to be opened if SFRCS signals are blocked.

Logic modifications emphasized the following features:

1) The valves listed above will close on an SFRCS trip.

2) These valves cannot be opened following an SFRCS trip without first blocking the SFRCS signal to them.

3) Momentary contact block switches were provided to override the SFRCS signal to these valves.

4) Blocking will be functional only when an SFRCS trip is present.

The block signal is automatically removed when the SFRCS trip resets so that additional SFRCS trips will reclose the valves.

To control steam generator pressure after an SFRCS actuation SFRCS signals to the atmospheric vent valves, ICS11A and ICS11B, wete blocked.

This modification emphasized the previous logic features in addition to the following:

1) The control of atmospheric vent valves can be handled by the Integrated Control System.

2)

Instrument air will be required for the operation of these valves.

Work was completed August 9, 1982.

FCR 79-221 Page 2 of 2 REASON FOR CllANCE: This FCR was implemented to provide a way to feed the steam generators with the SUFP to remove core decay heat in the event that both the Main Feed Pumps and the Auxiliary Feed Pumps are lost. This change will also provide a capability for steam generator pressure control af ter an SFRCS actuation.

SAFETY EVALUATION

SUMMARY

The purpose of the SFRCS is to protect the nuclear steam supply system by automatically initiating the Auxiliary Feedwater System in the event of one or more of the following incidents:

- a main feedwater line rupture

- a main steam line rupture

- a loss of both Main Feed Pumps

- a loss of four Reactor Coolant Pumps The changes made by this FCR will enhance the operation of the SFRCS.

Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REOUEST FCR No:

79-311 SYSTEM: Safety Features Actuation System COMPONENT: Various solenoid valves CHANCE TEST OR EXPERIMENT: FCR 79-311 allowed for the replacement of various solenoid valves associated with the Safety Features Actuation System (SFAS). The following valves were replaced under this FCR:

Valve No.

Function SV-229B Pressurizer Quench Tank Outlet Isolation Valve SV-235B Pressurizer Quench Tank Sample Isolation Valve SV-1719A Containment Ventilation Header Isolation Valve SV-1773A Reactor Coolant Drain Tank Header Isolation Valve SV-5006 Containment Purge Inlet Isolation Valve l

SV-5007 Containment Purge Outlet Isolation Valve SV-6831A Domineralized Water to Containment Isolation Valve Work associated with the replacement of these valves was completed November 30, 1981.

REASON FOR CHANCE: Upon review of all environmentally qualified solenoid valves provided by the Automatic Switch Company (ASCO), the above valves were found to have no qualification data. This data was not available when the original valves were purchased. Therefore, to have the proper documentation of these SFAS operated solenoid valves, the valves were l

replaced with duplicate ASCO valves with the proper documentation provided to Toledo Edison.

SAFETY EVALUATION

SUMMARY

The purpose of the solenoid valves listed above is to isolate Containment upon an SFAS actuation. The safety function of the SFAS is to automatically prevent or limit fission products and energy release from the core, to isolate the containment vessel, and to initiate operation of the Engineered Safety Features equipment in the event of a Loss of Coolant Accident (LOCA). By replacing the solenoid valves, the safety function of SFAS was not adversely affected. There-fore, an unreviewed safety question does not exist.

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l COMPLETED FACILITY CHANGE REQUEST l

l FCR NO: 80-074 l

SYSTEM: Auxiliary Feedwater COMPONENT: PSL-4930A&B and PSL-4931A&B CHANGE, TEST OR EXPERTMENT: FCR 80-074 was implemented to increase the time delay to close the steam inlet valves to the Auxiliary Feed Pump Turbines (AFPT) when there is low pressure in the suction lines of the Auxiliary Feed Pumps (AFPs). Valves HV-106 and HV-106A (for AFPT 1) and HV-107 and HV-107A (for AFPT 2) had a time delay of 0.5 seconds. The time delay has been increased to 2.5 seconds. This FCR also reconnected pressure switches PSL-4930A, 4930B, 4931A, and 4931B.

Each pressure switch was connected to each of the four steam inlet valves to the AFPTs.

The switches were connected as listed:

Switch No.

Valve PSL-4930A MS-106 PSL-4930B MS-106A PSL-4931A MS-107 PSL-4931B MS-107A These pressure switches will cause the steam inlet valves to close at 1 psig. Work was completed December 18, 1980.

REASON FOR CHANCE:

Prior to this FCR the steam inlet valves to the AFPTs have closed incorrectly due to low pressure dips in the AFP suction piping. The increasing of the time delay will prevent the spurious closure of the valves. The reconnection of the pressure switches to the steam inlet valves was performed to ensure a power supply failure would not prevent the valves to both the AFPTs from closing. Under the new connection, the power supply to HV-106 and HV-107 could fail with HV-106A and IIV-107A being maintained.

SAFETY EVALUATION

SUMMARY

The safety function of the Auxiliary Feedwater System is to provide feedwater to the steam generators for the removal of decay heat in the absence of main feedwater and to promote natural circu-lation in the Reactor Coolant System on a loss of all four Reactor Coolant Pumps. The changes made by this FCR do not reduce the safety function of this system. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REOUEST l

l FCR NO: 80-093 SYSTEM: Service Water COMPONENT: Various pipe supports and anchors CHANGE, TEST OR EXPERIMENT:

FCR 80-093 was implemented to modify various pipe anchors and supports needed to maintain the integrity of the Service Water System. Following is a list of the modified components. Work involved with this FCR was completed August 26, 1983.

Anchors A-135 A-136 A-137 A-138 A-139 A-140 A-141 A-144 A-161 A-217 A-262 A-370 Snubbers 41-HBC-34-H38 41-HBC-36-illi 41-HBC-43-H6 41-llBC-46-H44 41-IIBC-35-il36 41-11BC-36-1112 41-IIBC-45-H1 41-HBC-46-H45 41-HBC-35-H48 41-HBC-37-H2O 41-ilBC-45-H40 41-HBC-47-Il26 41-IIBC-36-II4 41-ilBC-37-H26 41-11BC-45-1143 41-ilBC-47-H29 41-HBC-36-il5 41-IIBC-37-H37 41-IIBC-45-H44 41-HBC-47-1132 41-ilBC-36-il6 41-HBC-42-Il2 41-ilBC-46-H38 41-llBC-47-H35 41-HBC-36-H7 41-11BC-42-119 41-ilBC-46-H41 41-HBC-47-Il38 41-11BC-36-119 41-11BC-42-1112 41-ilBC-46-H42 41-11BC-47-1147 41-HBC-36-il10 41-11BC-43-115 41-11BC-46-1143 41-HBC-47-H49 41-HCB-47-Il52 41-HCB-96-114 41-11CB-194-113 41-llBD-96-ll5 41-ilBD-98-H3 41-ilBD-191-ill 41-HBD-194-il4 REASON FOR CHANGE: The pipe supports and anchors modified were required as a result of re-analysis of the anchor / support in accordance with IE Bulletins 79-02 and/or 79-14.

SAFETY EVALUATION

SUMMARY

t The safety function of Service Water is to supply cooling water to the component cooling heat exchangers, the con-tainment air coolers, and the turbine plant cooling water heat exchangers during normal operation. By upgrading the listed Service Water System pipe anchors and supports, the safety function of the system is not decreased but enhanced. Therefore, an unreviewed safety question does not exist.

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i COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-119 SYSTEM: Main Steam and Core Flood COMPONENT: MS-106, MS-106A, MS-107 MS-107A, CF01A, and CF01B I

CHANGE TEST OR EXPERIMENT:

FCR 82-119 reset the torque switch settings for MS-106, MS-106A, MS-107, MS-107A, CFIA, and CFIB. The new torque switch settings are 2.5 for the open dial setting and 1.5 for the close dial setting. Work was completed November 28, 1984.

REASON FOR CHANCE: This FCR is to implement the torque switch settings specified in the Torrey Pines Technology Report to improve the operation of the valves listed above.

SAFETY EVALUATION

SUMMARY

The safety function of MS-106, MS-106A, MS-107, and MS-107A is to provide steam to or isolate steam to the Auxili-ary Feedwater Pump Turbine on a Steam and Feedwater Rupture Control System trip. The core flood valves, CFIA and CFIB, are isolation valves that isolate the Core Flood Tanks from that system's piping. The new settings will enhance the operation of these valves. Therefore, an unreviewed safety question does not exist.

COMPLETED FACILITY CHANGE REQUEST FCR NO:

84-108 SYSTEM:

Safety Features Actuation System Steam and Feedwater Rupture Control System COMPONENT: Various logic drawings l

CRANCE, TEST OR EXPERIMENT: This FCR updated logic drawings E-16, E-17 E-18. E-19, M-050, and M-051.

These drawings are associated with the Safety Features Actuation System (SFAS) and the Steam and Feedwater Rupture Control System (SFRCS). Work was completed December 4,1984.

REASON FOR CHANCE:

These drawings are being updated to represent the l

actual, correct plant conditions. Deficiencies were found to exist in these drawings following a review performed on Facility Change Requests in I

an effort to have all drawings accurately updated.

SAFETY EVALUATION

SUMMARY

The safety function of the SFAS is to automat-ically prevent or limit fission products and energy release from the core, i

to isolate the containment vessel, and to initiate operation of the Engineered Safety Features equipment in the event of a Loss of Coolant Accident.

The safety function of the SFRCS is to automatically initiate the Auxiliary Feedwater System in the event of a main feedwater line rupture, a main steam line rupture, the loss of both main feedwater pumps, or the loss of all four reactor coolant pumps.

This FCR will not decrease the safety function of either safety system.

Therefore, an unreviewed safety question does not exist.

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c COMPLETED FACILITY CilANGE REQUEST FCR NO: 84-134 SYSTEM: Reactor Coolant System COMPONENT: Cold leg nozzle plugs CHANGE, TEST OR EXPERIMENT: This FCR was implemented to perform a 10CFR50.59 review for the use of cold leg nozzle plugs during the 1984 Refueling Outage while performing maintenance activities within the Reactor Coolant System (RCS). Work involved with this FCR was completed January 11, 1985.

REASON FOR CllANCE: The use of the cold leg plugs and their possibic failure could drain the refueling canal and the upper portion of the reactor vessel with the potential loss of decay heat removal. Therefore, a 10CFR50.59 review was necessary prior to using the cold Icg nozzle plugs.

SAFETY EVALUATION

SUMMARY

The safety function of maintaining proper water levels in the refueling canal is to provide adequate shielding from and cooling for the fuel assemblies within the reactor vessel and while those assemblies are being moved during Mode 6.

No movement of fuel was permitted during these maintenance activities, and the pressure boundry for the section of the cole leg below the high pressure injection nozzle was kept intact. Therefore, based on these observations, the use of the RCS cold leg nnzzle plugs did not present an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-174 SYSTEM:

Fire Barriers I

COMPONENT:

One-hour rated fire barrier over conduit 36591A CHANCE, TEST OR EXPERIMENT:

FCR 84-174 installed a temporary one-hour rated fire barrier over conduit 36591A in Room 322. This barrier is to remain in place until a permanent modification which complies with 10CFR50, Appendix R is implemented. This conduit contains cables required for Reactor Coolant System pressure and temperature and steam generator pressure and icvel instrumentation. Work involved with FCR 84-174 was I

completed December 23, 1984.

REASON FOR CHANCE: This. change is the result of the Toledo Edison Appen-dix R non-compliance " Seriousness Assessment" meeting held September 24, 1984.

In this meeting, it was determined that conduit 36591A should be protected with a temporary one-hour fire barrier until a permanent modifi-cation is implemented.

l SAFETY EVALUATION

SUMMARY

The safety function of the cables contained in l

the conduit is to provide an indication of several plant parameters at the auxiliary shutdown panel. Therefore, the addition of a temporary fire l

barrier will maintain system integrity until a permanent one-hour fire l

barrier can be installed. Therefore, an unreviewed safety question does l

not exist.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-194 SYSTEMt Reactor Vessel Internal and Lifting Facilities COMPONENT:

Fuel Assembly, S/N NJ91CR CHANCE. TEST OR EXPERIMENT FCR 84-194 replaced the fuel assembly holddown spring on fuel assembly S/N NJ91CR. The original holddown spring was of the B6 type. This was replaced with the B7 type holddown assembly. Work was completed February 13, 1985.

REASON FOR CHANGE:.The holddown spring for assembly S/N NJ91CR was found broAen during an inspection of the fuel assembly during the 1984 Refueling Outage.

SAFETY EVALUATION

SUMMARY

The safety function of the holddown spring is to hold down the fuel assembly and to allow for differential expansion between the fuel assembly and the reactor vessel internals. Originally, the holddown spring was type B6 which was constructed of Inconel X-750.

The new holddown spring is type B7 which is fabricated of Inconel 718.

Because of the material it is composed of, use of the type B7 holddown spring will provide approximately a 20% higher yield stress. Replacement of the type B6 fuel assembly holddown spring with a type B7 spring will reduce the potential for breakage of the spring due to its higher fatigue capacity. Therefore, since the safety function of the holddown spring is not reduced, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-218 SYSTEM: Startup Feed Pump and Auxiliaries System (045-02)

COMPONENT:

FW-32 l

CHANGE, TEST OR EXPERIMENT: FCR 84-218 modified drawing 7749-M-006B to show valve FW-32, the suction from the Deaerator Storage Tank, normally closed. This drawing change was completed February 8,1985.

REASON FOR CHANGE: This change was made to the drawing so valve FW-32 could be shown in its normal, correct position per the Updated Safety Analysis Report (USAR).

SAFETY EVALUATION

SUMMARY

The safety function of FW-32 on drawing 7749-M-006B is to isolate the startup feed pump suction line from the Deaerator Storage Tank before it enters the Auxiliary Feedvater Pump (AFP)

Rooms 237 and 238. This eliminates the moderate energy concerns and their potential effects on the AFP and associated components of the Auxiliary Feedwater System. The moderate energy concern is introduced when FW-32 is opened for a limited amount of time during plant startup. This valve is closed any other time. Therefore, since the startup feed pump system is not in operation for most of the plant operating time, the probability of introducing moderate energy effects into Rooms 237 and 238 are greatly reduced.

The drawing modification to show FW-32 normally closed on drawing 7749-M-006B does not present an unreviewed safety question since FW-32 in a normally closed position drastically decreases the time period that the AFP rooms are subjected to moderate energy concerns and, therefore, greatly reducen l

the probability of introducing moderate energy effects into AFP Rooms 237 and 238.

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e COMPLETED FACILITY CHANCE REQUEST FCR NO: 85-004 SYSTEM: Reactor Coolant System COMPONENT: Pressurizer CHANCE, TEST OR EXPERIMENT: This FCR revised drawing 7749-M-190-HBW-135-6 to reflect the correct, actual condition of the plant. The drawing revision involved correcting the drawing to show the placement of shims behind a baseplate on support PS-H36, which is associated with the pres-surizer. Work was completed April 1, 1985.

REASON FOR CllANGE: This change was the result of Non-conformance Report 84-178. The non-conformance report was initiated after the performance of Surveillance Test ST 5044.01 Inspection of Safety Related liydraulic Snubbers. During this test, the baseplate for PS-ll36 was 1/4 inch away from the wall. To correct this, shims were added behind a baseplate on support PS-H36. The drawing was updated to reflect the correct, actuni condition of the plant.

SAFETY EVALUATION

SUMMARY

The safety function of the preasurizer is to maintain pressure in the Reactor Coolant System during all phases of plant operation. This FCR did not create any adverse environment to the pres-nurizer, but instead it represents the actual condition of the plant.

Therefore, an unreviewed safety question does not exist.

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TOLEDO

%me EDISON August 9, 1985 Log No. K85-1193 Files RR 2 (P-6-85-07)

Docket No. 50-346 License No. NPF-3 Mr. Norman llaller, Director Office of Management and Program Analysis U. S. Nuc1 car Regulatory Commission Washington, D.C.

20555

Dear Mr. Haller:

Monthly Operating Report, July 1985 Davis-Benne Nucicar Power Station Unit 1 Enclosed are ten copics of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of July 1985.

If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000 Extension 290.

Yours truly, J

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Louis F. Storz Plant Manager Davin-Besse Nuclear Power Station LFS/MK/1jk Enclosures cca Mr. James C. Keppler, w/1 Regional Administrator, Region III Mr. James H. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Walt Rogers, w/l NRC Resident Inspector

1. l LJK/002 THE TOLEDO E0!GON COMPANY LOlGON PLAZA 300 MA0iSON AVENut Tutt00, Oh10 43052 i

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