ML20212J255

Monthly Operating Rept for Jul 1986

ML20212J255
Person / Time
Site:	Davis Besse
Issue date:	07/31/1986
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
KB86-0683, KB86-683, NUDOCS 8608140277
Download: ML20212J255 (15)

Text

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o AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 Davis-Besse Unit 1 UNIT DATE August 8, 1986 COMPLETED BY Morteza Khazrai (419) 249-5000, TELEPHONE Ext. 7290 MONT1f July 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 0 g7 0

O 3g 0

3 0

19 0

4 0

20 0

5 0

0 21 6

0 22 0

7 0

n 0

8 0

24 0

9 0

25 0

10 0

26 0

g3 0

27 0

12 0

0 28 13 0

0 29 14 0

0 30 15 0

33 0

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.

(9/771 L

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE August 6, 1986 COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000, Ext. 7290 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

2. Reporting Period:

July 1986

3. Licensed Thermal Power (MWr):

2772

4. Nameplate Rating (Gross MWe):

925

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 (Irems Number 3 Through 7) Since Last Report. Give Reasons:

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

10. Reasons For Restrictions,If Any:

i This Month Yr.-to.Date Cumulative i1. Hours In Reporting Period 744 5,087 70,152

12. Number Of Hours Reactor Was Critical 0.0 0.0 35.877.1

13. Reactor Reserve Shutdown Hours 0.0 0.0 4.058.8

14. Hours Generator On.Line 0.0 0.0 34,371.8

15. Unit Reserve Shutdown Hours 0.0 0.0 1 732.5

16. Gross Thermal Energy Generated (MWH) 0.0 0.0 81.297.600

17. Gross Electrical Energy Generated (MWH) 0.,0_.

0.0 26.933.622

18. Net Electrical Energy Generated (MWH) 0.0 0.0 25.233.177

19. Unit Service Factor 0.0 0.0 49

20. Unit Availability Facts.r 0.0 0.0 51.5

21. Unit Capacity Factor (Usir.g MDC Net) 0.0 0.0 41.8

22. Unit Capacity Factor (Using DER Net) 0.0 0.0 39.7 1

23. Unit Forced Outage Rate 100.0 100.0 32.8

24. Shutdowns Scheduled Over Next 6 Months (Type. Date,and Duration of Each1:

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

October 18, 1986

26. Units In Test Status (Prior to Commercial Operation):

Forecast Achiesed INITIAL CRITICALITY

~

INITIAL ELECTRICITY COMMERCIAL OPERATION (9/771

DOCKET NO.

50-346 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit 1 DATE August 8, 1986 COMPLETED BY Morteza Khazrai REPORT MONTH July 1986 TELEPHONE (419) 249-5000 Ext. 7290 "u

E g

g "g

Licensee ag g

Cause & Corrective Event g

gg g

guy go gg Action to No.

Date gg y

gy Report #

gu aa Prevent Racurrence zmf av a

o 7

85 06 09 F-744 A

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

See Operational Sunmary for further details.

1 2

F Fo m d 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 5

G-Operational Error (Explain) 9-Other (Explain)

Exhibit I - Same Source (9/77)

H-Other (Explain)

OPERATIONAL

SUMMARY

JULY 1986 The unit remained shutdown the entire month of July following the reactor trip on June 9, 1985. Corrective actions and system upgrades continue.

Below are some of the major activities performed during this month:

1)

Continued testing as part of the System Review and Test Program.

2)

Continued Motor Operated Valves Analysis Test (MOVATS) activities.

3)

Continued Raychem investigation and followup corrective actions.

4)

Completed required Environmental Qualification (EQ) field work to satisfy the EQ audit requirements.

5)

Initiated work for replacement of remaining three Reactor Coolant Pumps rotating assemblies.

6)

Completed mechanical / electrical field work for Service Water Pump #1 - awaiting final Quality Control approval prior to returning the unit to service.

7)

Provide management support to Maintenance and Operations for Emergency Operations Plan (EOP) activities.

8)

Completed field work in preparation of Integrated Safety Features Actuation System (SFAS) testing continues.

l l

l r

l t

REFUELING INFORMATION DATE: July 1986 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown: October, 1987 3.

Scheduled date for restart following refueling: December, 1987 4.

Will refueling or resumption of operction 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)?

I Ans: Expect the R61oad Report to require standard reload fuel design Technical Specification changes (3/4.1 Reactivity Centrol Systens and 3/4.2 Power Distribution Limits).

5.

Scheduled date(s) for submitting proposed licensing action and supporting information:

Summer, 1987 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) 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:

1995 - assuming ability to unload the entire core into the spent fuel pool is maintained.

i BMS/005

_ _ ____,_. _ _. _ _ _ _ _, _ ~. - _ _ _

COMPLETED FACILITY CHANGE REQUESTS FCR NO.83-057 SYSTEM Main Steam COMPONENT ZSICS 11A, ZSICS 11B, ZS100-1, ZS101-1, ZS375 and ZS394 limit switches CHANGE, TEST OR EXPERIMENT This FCR replaced the following existing limit switches with environmen-tally qualified NAMCO aeries EA-180 limit switches ZSICS 11A ZSICS 11B ZS100-1 ZS101-1 ZS375 ZS394 This FCR was closed July 4, 1986 REASON FOR CHANGE These limit switches were replaced to meet the requirements of NRC Bulletin 79-01B on Environmental Qualification of Safety Related Equipment.

SAFEIT EVALUATION SUMMA 2Y This FCR replaced various limit switches with environmentally qualified limit switches. The safety function of these associated valves is con-tainment isolation. The limit switches are for valve position indication only and have no control function. Since the FCR involves only replace-ment of existing limit switches with qualified models these limit switches will still perform their intended safety function after these changes are made.

The changes proposed by this FCR do not involve any unreviewed safety question.

DD C/153

COMPLETED FACILITY CHANGE REQUESTS FCR 83-058 SERVICE Various COMPONENT Junction Boxes CHANGE, TEST OR EXPERIMENT This FCR allowed 1 2/3" diameter holes to be drilled on the following junction boxes; JT-2917 JT-3953, JT-3954, P2C5G1, PIC2L1, P1P3B1, P2P4C1, P2 PSF 1, PlP2M1, PlLlL1, P2L4G1, P3L451, P4LIG1, PIC551 and P2C5C1.

This FCR was closed January 15, 1986 REASON FOR CHANGE The installed junction boxes could not withstand the pressure rise due to a postulated pipe rupture. Drilling the holes in the junction boxes reduced the pressure differential to acceptable levels.

SAFETY EVALUATION

SUMMARY

1 The junction boxes were originally installed to house terminal blocks which were not environmentally qualified. Since then the terminal blocks have been removed and environmentally-qualified Raychem shrinkable tubes were installed over the splices. The junction boxes cannot withstand the pressure rise due to a postulated pipe rupture. Drilling holes reduced the pressure differcutial across the cover to acceptable levels, thereby preventing any damage to electrical penetration and junction devices.

The basis for drilling the holes is per EDS Nuclear Study on Environmental Qualification of Safety Related Electrical Equipment at Davis-Besse Nuclear Plant. Based on the above summary no new unreviewed safety question exists.

l DD C/153

COMPLETED FACILITY CHANGE REQUESTS FCR NO.

84-0196 SYSTEM Service Water COMPONENT MV-2929, MV-2930, MV-2931 and MV-2932 CHANGE, TEST OR EXPERIMENT This FCR changed the torque setting to the following valves as follows; EXISTING SETTTMS NEW SETTINGS OPEN CLOSE OPEN CLOSE MV-2929 2.5 2.5 3.0 1.0 MV-2930 1.75 1.75 3.0 1.0 MV-2931 1.75 1.75 3.0 1.0 MV-2932 1.75 1.75 3.0 1.0 This FCR was closed June 23, 1986 REASON FOR CHANGE The proposed torque settings were based on Torrey Pines Technology Report on Limitorque Motor Operated Valves. The new settings are to improve valve reliability.

SAFETY EVALUATION

SUMMARY

1.

The safety function of the torque switch for butterfly valves is to close or open the valve if the limit switch fails to perform its function.

2.

The second safety function of the torque switch is to break the circuit in case of high mechanical force to prevent any damage to the valve.

The proposed settings are recommendations from Torrey Pines Technology Report on Limitorque Motor Operated Valves (Report No. 23981A). The calculation indicates that reducing the close torque switch settings still will allow the proper closing of the valves, if the limit switch fails to perform its function.

The proposed settings do not affect the safety function of the torque switch. The new settings improved the reliability of these valves.

The work authorized by this FCR did not create any new adverse environment and does not constitute an unreviewed safety question.

DD C/153

COMPLETED FACILITY CHANGE REQUESTS FCR NO 83-052 SYSTEM Main Steam COMPONENT SV 275, SV 394, SV101A, SV1D1B, SV101C, SV101D, SV101E, SV100A, SV100B, SV100C, SV100D and SV100E CHANGE, TEST OR EXPERIMENT This FCR replaced the above solenoid valves to meet the requirements of NRC Bulletin 79-01B.

This FCR was closed July 4, 1986 REASON FOR CHANGE These solenoid valves had to be replaced to meet the requirements of NRC Bulletin 70-01B on Environmental Qualification of Safety Related Equipment. The deadline for replacement of equipment which are not quali-fled is March 31, 1985.

SAFETY EVALUATION

SUMMARY

This FCR replaced various solenoid valves with environmentally qualified valves. The original solenoid valves did not meet Environmental Qualifi-cation Requirements. Since this FCR only involved model changes, the solenoid valves will still perform their intended safety functions after these changes were made.

Therefore, the changes proposed by this FCR do not involve any unreviewed safety question.

a DD C/153

COMPLETED FACILITY CHANGE REQUESTS FCR NO 82-168 SYSTEM Decay Heat and Low Pressure Injection COMPONENT PSH RC2B4 CHANGE, TEST OR EXPERIMENT This FCR replaced PSH RC2B4 Switch with a Environmental Static-0-Ring Pressure Switch capable of withstanding normal Reactor Coolant System operating pressure.

This FCR was closed February 22, 1986 REASON FOR CHANGE The pressure switch (PSH RC2B4) was replaced with a Static-0-Ring Pressure Switch to meet the requirements of NRC Bulletin 79-01B on Environmental Qualification of Safety Related Equipment.

SAFETY EVALUATION

SUMMARY

The replacement of the existing nonqualified switch with an Environmental-ly Qualified Static-0-Ring Pressure Switch because it will operate reliability in a harsh environment. A different range and a smaller deadband on the new switch will eliminate the problem which the existing switch had when resetting on decreasing RCS pressure. The change proposed by FCR 82-168 does not involve any unreviewed safety question.

DD C/153

4 COMPLETED FACILITY CHANGE REQUESTS FCR NO 80-255 SYSTEM SFRCS COMPONENT Channel 4 CHANGE, TEST OR EXPERIMENT This FCR modified the 15 VDC power supply for SFRCS Channel 4 as follows.

1.

Readjusted the output voltage to 14.5 VDC 2.

Readjusted the overvoltage protection to 17 VDC.

This FCR was closed May 30, 1986 REASON FOR CHANGE This FCR lowered the output voltage and raised the overvoltage protection to prevent spurious voltage spikes from tripping the power supply. This was a temporary modification while awaiting further evaluation by CCC/Sorensen.

SAFETY EVALUATION

SUMMARY

The safety function of the over voltage protection is to protect the system components from exposure to a voltage high enough to damage the components. The system is capable of operation to 18 VDC, therefore an over voltage trip set at 17 VDC will not effect the safety function.

The safety function of the operating voltage is to provide power to the system components. The system can operate at a voltage as low as 13 VDC, therefore by reducing the operating voltage to 14.5 VDC the safety func-tion will not be affected.

Based on the above information, no unreviewed safety question exists.

DD C/153

COMPLETED FACILITY CHANGE REQUEST FCR NO 85-149 SYSTEM Steam Generators COMPONENT OTSG E24-1 and E24-2 CHANCE, TEST OR EXPERIMENT This FCR performed a 10CFR50.59 review to help determine the effects of the June 9, 1985, transient on the integrity of the Once Thru Steam Generators.

This FCR was closed October 11, 1985.

REASON FOR CHANGE This FCR demonstrated that the June 9, 1985, transient had no adverse effects on the structural integrity of both steam generators and their associated components.

SAFETY EVALUATION

SUMMARY

This FCR performed a 10CFR50.59 review of the effects of the June 9, 1985, event on the integrity of the Once Thru Steam Generators (OTSG) at Davis-Besse. Babcock & Wilcox was requested to perform an evaluation of the Davis-Besse OTSGS.

In summary, the following analyses were performed:

1.

Main feedwater nozzle thermal shock 2.

Auxiliary feedwater nozzle thermal shock 3.

Thermal shock of OTSG Tubes due to impingement of auxiliary feedwater 4.

Axial compressive in tubes due to shell to tube temperature differential 5.

Thermal shock on lower tube sheet due to main feedwater flow The results of this evaluation show that the June 9, 1985, transient had no adverse structural effect on the steam generators.

Pursuant to the above, since all of the components of the OTSG have been found acceptable, it is concluded that the June 9, 1985, loss of feedwater transient had no adverse structural effects on the OTSGS. Babcock &

Wilcox recommended no additional testing of the OTSC components prior to the restart of the plant. However, it is recommended that the outer tubes be inspected during the next scheduled Eddy current inspection of OTSG 2 for any indication of wear near the fifteenth tube support plate.

Based on the above summary it is concluded no unreviewed safety question exists.

DD C/153

COMPLETED FACILITY CHANGE REQUEST FCR NO 78-297 SYSTEM Reactor Coolant System COMPONENT RCP Neutron Power Interlock CHANGE, TEST OR EXPERIMENT This FCR requested approval of testing to obtain operational data on the restarting of the 4th RCP above 22% power to confirm the simulator results obtained by Babcock & Wilcox.

This FCR was closed August 18, 1983.

REASON FOR CHANGE To verify test data with the simulator results in order to be able to raise the Neutron Power Interlock setpoint to allow the 4th RCP to be restarted at a higher power level than the present 22% interlock setpoint.

SAFETY EVALUATION

SUMMARY

The 22% full power ICS interlock setpoint ensured that a pressure trip setpoint (high or low) would not be reached before reaching equilibrium conditions after the starting of the fourth RCP. An analysis of the data provided by the B&W simulator at 50% full power revealed that the highest and lowest pressure after starting the fourth RCP was 2192 psig and 2090 psig, which arc veil within the pressure trip setpoint.

The 22% interlock is an non-nuclear safety related interlock and was introduced in tle original integrated control system (ICS) to avoid the possibility of frequent R.C. pressure trips.

Based on the above, starting a fourth RCP at 50% full power will not compromise nuclear safety and does not involve an unreviewed safety question.

DD C/153

o COMPLETED FACILITY CHANGE REQUEST FCR NO 79-088 SYSTEM Reactor COMPONENT Core GANGE, TEST OR EXPERIMENT 1.

This FCR replaced 44 fuel assemblies for the first refueling.

2.

Inserted 2 regenerative neutron sources into fuel assemblies as specified in BAW-1598 reload report.

This FCR was closed June 11, 1986.

REASON FOR CHANGE 1.

The change of 44 fuel assemblies instead of 52 fuel assemblies was desirable because of economic consideration.

2.

The insertion of 2 regenerative neutron sources was necessary for continued operations.

SAFETY EVALUATION

SUMMARY

The original reload report for cycle 2 was performed based on 248 EFPD, since then the projected cycle length was updated by B&W (Ref. 1 & 2) to 272 EFPD and subsequently 285 EFPD without APSR withdrawal. The cycle length was further extended (Ref. 3) to 306 EFPD assuming APSR withdrawal beyond 250 EFPD. This FCR provided the justification for continued cycle 2 operation to 285 EFPD without APSR withdrawal in case it is needed to operate in this arrangement.

Pursuant to the above, it is concluded that there is no unreviewed safety question involved with the proposed changes.

DD C/153

a TOLEDO

%ss EDISON August 8, 1986 Log No. KB86-0683 File: RR 2 (P-6-86-07)

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

20555

Dear Mr. Haller:

Monthly Operating Report, July 1986 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 1986.

If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000 Extension 7290.

Yours truly, L h$

Louis F. Storz Plant Manager DavJs-Besse Nuclear Power Station LFS/MK/ljk Enclosures l

cc:

Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/1 NRC Resident Inspector l

l LJE/002 g

THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652

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