ML20197D035

Monthly Operating Rept for Feb 1986

ML20197D035
Person / Time
Site:	Davis Besse
Issue date:	02/28/1986
From:	Khazrdi M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
NUDOCS 8605140106
Download: ML20197D035 (21)

Text

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AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

UNIT Davis-Besse Unit 1 DATE March 10, 1986 COMPLETED BY Morteza Khazrai TELEPHONE 419-249-5000 Ext. 7290 MO G February 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1 0

g7 0

2 0

is 0

3 0

39 0

4 0

20 0

5 0

0 21 6

0 2:

0 7

0 0

3 3

0 24 0

9 0

3 0

10 0

26 0

O 0

II 27 12 0

0

g 13 0

29 14 0

30 15 0

3 16 0

INSTRUCTIONS On this format, list the average dady unit power leselin MWe Net for each day in the reporting month. Compute to the nearest whole rnegawatt.

(9/77) 8605140106 860228 PDR ADOCK 05000346 l l R

PDR

r OPERATING DATA REPORT DOCKET NO. ' 50-346 DATE March 10, 1986 COMPLETED BY Morteza Khazral TELEPHONE 419-249-$000, Ext. 7290 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

2. Reportmg Pened.

February 1986

3. Lacensed Thermal Power (MWr):

2772

4. Nameplate Rating (Gross MWe):

925

5. Desagn Electrical Rating (Net MWe):

906

6. Maximum Dependable Capacity (Gross MWe):

904

7. Maximum Dependable Capacity (Net alwel:

860

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

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

10. Reasons For Restrictions,if Any:

This Month Yr.-to.Date Cumulative

11. Hours In Reporting Period 672.0 1,416.0 66,481.0

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

13. Reactor Reserve Stutdown Hours 0.0 0.0 4.058,8 I4. Hours Generator On-Line 0.0 0.0 34.371.8

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Grom Thermal Energy Generated (MWH) 0.0 0.0 81,297,600

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

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 51.7

20. Unit Availability Factor 0.0 0.0 54.3

21. Unit Capacity Factor (Using MDC Net) 0.0 0.0 44.1

22. Unit Capacity FactoriUsing DER Net) 0.0 0.0 41.9

23. Unit Forced Outage Rate 100.0 100.0 27.6

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

1

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

April 28, 1986

]

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

Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (4/77)

DOCKET No.

50-346 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit I

/

DATE March 10, 1986 COMPLETED BY Morteza Khazrai REPORT MONTil February 1986 TELEPHONE 419-249-5000. Ext. 7290 m

W O

M 8m "a

NU Licensee a.

E.n Cause & Corrective

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'8 U $

Event 3$

Action to No.

Date O$

'8 $ "

Report #

$0 0

Prevent Recurrence S6 E $! $

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85 06 09 F

672 A

4 LER 85-013 JK SC The unit remained shutdown follow-Con't ing the reactor trip on June 9, 1986.

See Operational Summary for further

details, a

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2 3

4 F: Forced Reason:

Method:

Exhibit G - Instructions 1

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

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

Exh'_ bit I - Same Source

]

(9/77)

H-Other (Explain) t

OPERATIONAL

SUMMARY

FEBRUARY, 1986 The unit remained shutdown the entire month of February following the reactor trip on June 9, 1986. Investigation of the causes of the event and corrective actions continue. See NUREG 1154 for further details.

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

1)

Continued MOVATS testing 2)

Completed FCRs on #1 Emergency Diesel Generator 3)

Completed Channels #1 and #2 of SFAS wiring i

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1 REFUELING INFORMATION DATE: February, 1986 i

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1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown: March 1, 1987 3.

Scheduled date for restart following refueling: May 10, 1987 1

4.

Will refueling or resumption of operation thereafter require a 1

technical specification change or other license amendment? If answer l

is yes, what in general will these be? If answer is no, har the reload fuel design and core configuration been reviewed by your Plant 4

l 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 Reload Report to require standard reload fuel desian 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, 1986 6.

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

j new or different fuel design or suppliec, unreviewed design or performance analysis methods, significant changes in fuel design, new-i operating procedures.

Ans: None identified to date.

The number of fuel assemblies (a) in the core and (b) in the spent l

7.

fuel storage pool.

a i

(a) 177 (b) 204 - Spent Fuel Assemblies 8.

The present licensed spent fuel pool storage capacity and the size of j

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 - assumina ability to unload the entire core into the 4

spent fuel pool is maintained.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-139 SYSTEM: Miscellaneous COMPONENT: N/A CHANGE, TEST OR EXPERIMENT: This FCR was issued to perform the following tasks:

1)

Reissue miscellaneous Bechtel drawings previously issued as part of Davis-Besse construction specifications only 2)

Deletions of drawings 3)

To incorporate minor changes Work was completed October 14, 1983.

REASON FOR CHANGE: This change was necessary since all these drrwings are attachments to specifications, but which were never issued forma 113 as drawings.

SAFETY EVALUATION

SUMMARY

This safety evaluation is necessary since some of the reissued drawings are "Q".

The safety function of the "Q" design document is to record how a safety system was built. The reissue and the minor change of the "Q" drawings did not change the safety function, and no possibility of an adverse environment was created. It standardized the drawing control process, therefore, this is not an unreviewed safety question.

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COMPLETED FACILITY CRANGE REQUEST FCR NO: 83-143 SYSTEM: Control Rod Drive Control System (CRDCS)

I COMPONENT: Programmer Control Fuses F-3-1 and F-3-2 i

CHANGE, TEST OR EXPERIMENT: This FCR will allow for the removal of fuses F-3-1 and F-3-2 from the safety group programmer controllers for the CRDCS. These fuses are required for the movement of rods in Groups 1, 2, g

3, and 4.

These fuses are on the normal power supply. Work was completed February 9, 1984.

REASON FOR CHANGE: This change will prevent the inadvertent energizing of the silicon control rectifier power supplies which cause the control rod drive mechanisms to rotate on Groups 1, 2, 3, and 4.

This change will also prevent an inadvertent runback due to an asymmetric fault on Groups 1, 2, 3, and 4.

SAFETY EVALUATION

SUMMARY

The function of the programmer motor is to

.i provide sequential in-and-out commands to each group or rod as directed to insert or withdraw safety rods. Therefore, af ter safety tods are withdrawn, the fuses could be removed.

The Reactor Protection System (RPS) or Anticipatory Reactor Trip System (ARTS) are not dependent on safety groups being energized. This function being performed does not constitute an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-159 SYSTEM: Auxiliary Feedwater COMPONENT: MV-360 and MV-388 CHANGE, TEST OR EXPERIMENT: This FCR changed the torque settings of the Auxiliary Feed Pumps (AFP) discharge valves MV-360 (AFP 1-1) and MV-388 (AFP 1-2) from 2.75 open/ closed to 1.0 open/ closed. Work was completed January 15, 1985.

REASON FOR CHANGE: The new torque switch settings were made to improve valve reliability. The basis for the new settings is from the Torrey Pines Technology Report on Limitorque Motor Operated Valves.

SAFETY EVALUATION

SUMMARY

The safety function of the torque switch is to close a valve tightly enough to prevent any leakage and to break the circuit in case of high mechanical force to prevent any damage to the valve or to break the circuit in closing the valve to prevent overtraveling of the valve stem. The new switch settings will allow the valve to properly close, thus preventing any leakage or seat damage. The new open switch setting will prevent overtraveling of the valve stem. The new torque switch settings enhance the operation of these valves. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 85-002 SYSTEM: Containment Normal Sump COMPONENT: MV-2012B CHANGE, TEST OR EXPERIMENT: FCR 85-002 changed the open torque setting for the Containment Normal Sump Isolation Valve, MV-2012B, from 1.0 to 1.5.

Work was completed January 9,1985.

REASON FOR CHANGE: While inspecting MV-2012B during the 1984 Refueling Outage, the open torque switch setting was set at 1.0.

When the valve was checked for operability, it torqued out in the open direction. The torque setting was increased to 1.5, and the valve and operator opened properly.

SAFETY EVALUATION

SUMMARY

The safety function of the torque dial setting is to allow the Centainment Normal Sump Isolation Valve to open properly; therefore, preventing untraveling of the valve stem. The new dial setting will not degrade the safety function of the torque switch, but enhance the operation of the valve. Therefore, an unreviewed safety question does not exist.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-168 SYSTEM: Containment Spray COMPONENT: MV-1530 and MV-1531

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l CHANGE, TEST OR EXPERIMENT: This FCR changed the torque switch settings for the Containment Spray Pump i Header Valve, MV-1530, and the Containment Spray Pump 2 Outlet Valve, MV-1531, from 2.0 open to 1.0 open. Work was completed January 4, 1985.

REASON FOR CHANGE: These changes were made to improve valve reliability.

The basis for the new settings are from the Torrey Pines Technology Report on Limitorque Motor Operated Valves.

SAFETY EVALUATION

SUMMARY

The safety function of the torque switch is to close a valve tightly enough to prevent any leakage and to break the circuit in case of high mechanical force to prevent any damage to the valve or to break the circuit in closing the valve to prevent overtraveling of the valve stem. The new switch settings will allow the valve to close i

l properly, thus preventing any leakage or seat damage. The new open switch setting will prevent overtraveling of the valve stem. The new torque switch settings enhance the operation of these valves. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-273 SYSTEM: Reactor Protection System (RPS)

COMPONENT: N/A i

CHANGE, TEST OR EXPERIMENT: This FCR will modify the field wiring side of RPS channel location 1-10-6 Terminals 3-4 in Cabinets C5762, C5763, C5755, and C5756 by installing studs large enough to properly retain an alligator clip. This work was completed February 6,1986.

1 R,EASON FOR CHANGE: This modification was requested because an alligator j

clip slipped off of one of the terminals causing a short across the l

essential 120VAC instrument bus, which in turn tripped inverter YV4, causing deenergization of SFAS Channel 4 and RPS Channel 4.

This also caused the closure of DH-11 isolating the Decay Heat System.

SAFETY EVALUATION

SUMMARY

This FCR involves the installation of terminal i

block spacers to prevent equipment damage and personnel injury during the testing of the RPS channels. All modifications are internal to the RPS cabinets and will not create any new adverse environment. An unreviewed safety question does not exist.

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j COMPLETED FACILITY CHANGE REQUEST i

FCR NO: 79-383 I

SYSTEM: Reactor Coolant System j

COMPONENT: Flow Indicator Sensing Lines CHANCE TEST OR EXPERIMENT: This FCR will install redundant flow transmitter sensing lines for each reactor coolant loop from the second root calve to the flow transmitters. There will be two flow transmitters on each line.

Work was completed February 6, 1986.

i REASON FOR CHANGE: This modification was required by License Condition 2.C(3)(e) of the Davis-Besse Unit #1 Operating License.

SAFETY EVALUATION

SUMMARY

This modification is required by the License Condition 2.c(3)(e) of Davis-Besse Unit #1 Operating License. The possibility of an accident or a malfunction of a different type that is not bounded by l

the previous analysis in the FSAR has not been created. The margin of safety as defined in the basis for any Davis-Besse Unit #1 Technical Specification has not been decreased.

No unreviewed safety question is involved.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-306 SYSTEM: Radiation Monitoring System COMPONENT: RE2004, RE2005, RE2006, RE2007, RE2387, and RE2389 i

CHANGE, TEST OR EXPERIMENT: This FCR will change the location of the following radiation monitors:

Monitor Present Location New Location l

RE2004 A

A g RE2005 C

C y RE2006 D

D RE2007 E

B 1

RE2387 B

Same l

RE2389 F

Same 1

This work was completed January 23, 1986.

REASON FOR CHANGE: Failure of the Victoreen radiation monitors has been a j

continuing problem and as a result, personnel neutron exposures have been received due to inspection and repairs of the monitors during power operations. Exposures would be greatly reduced by relocation of these monitors.

SAFETY EVALUATION

SUMMARY

The relocation of SFAS radiation monitors RE2004, RE2005, RE2006, and RE2007 into Penetration Room Elevation 585' 1

and Elevation 603' will not affect the safety function of these monitors.

This location will increase the reliability of the monitors because the penetration room environment is lower in temperature and humidity than the Containment Vessel. Also, Station personnel radiation exposures will be i

lower.

This is not an unreviewed safety question.

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i COMPLETED FACILITY CHANGE REQUEST i

FCR NO: 83-066 SYSTEM: Station AC Electrical Power Distribution COMPONENT: Miscellaneous I

i CHANGE, TEST OR EXPERIMENT: This FCR implemented the testing of the Station Electrical Distribution System through the degraded bus voltage i

test.

j Work involved with this FCR was completed July 23, 1983.

I REASON FOR CHANGE: Testing was required to verify the assumptions of the j

analytical study performed to determine the adequacy of the electrical power system from the standpoint of operability of Class 1E equipment i

during a degraded grid voltage condition.

1 SAFETY EVALUATION SUMHARY: The defeat of the undervoltage relays would l

have a safety significance only if coincident with a LOCA which would i

demand loading of safety related equipment on the essential buses. IF a LOCA were not in progress, the undervoltage relay defeat would produce no 1

equipment degradation. Therefore, the actual unreliability imparted to safety related equipment caused by the use of the bypass would be less than 2 x 10 " multiplied by the likelihood of a LOCA. Given the small likelihood of a LOCA, the added unreliability is an acceptably small number.

I It is concluded, therefore, that this is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR No: 79-228 SYSTEM: Hydraulic Snubbers i

COMPONENT: A-399 CHANGE, TEST OR EXPERIMENT: This FCR will allow modification to change the remote reservoir to a local reservoir for anchor A-399.

Work was completed September 29, 1983.

REASON FOR CHANGE: This snubber is a vertical snubber with a remote reservoir. Problems have developed with air entering the snubber. By changing the remote reservoir to a local reservoir, air entering the snubber should be decreased significantly.

SAFETY EVALUATION SU)0MRY: This FCR involves modifying the remote reservoirs for snubbers on anchor A-399 to locally mounted reservoirs.

No new adverse environmental will be created by this seismic modification.

No further EIT involvement is required. Therefore, no unreviewed safety question exists.

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i COMPT.ETED FACILITY CHANGE REQUEST l

I FCR NO: 79-425 SYSTEM: Containment Vessel Pressure Measurement 1

COMPONENT: N/A

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CHANCE. TEST OR EXPERIMENT: This FCR involved the addition of pressure transmitters (PT-4587 and PT-4588) for wide range containment pressure monitoring. This work was completed February 9, 1986.

REASON FOR CHANGE: Pressure transmitters PT-4587 and PT-4588 vere installed j

in response to October 23, 1979 Toledo Edison letter Serial No. 546 to the j

NRC and NRC letters dated September 13, 1979 and October 30, 1979 on followup actions resulting from the NRC staff reviews regarding the Three

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Mile Island Unit #2 accident.

I SAFETY EVALUATION

SUMMARY

The proposed addition of a wide range containment vessel pressure measurement is nuclear safety related due to the design i

and installation criteria set by the NRC Criterion 13. Appendix A to 10CFR Part 50 and Qualified to Regulatory Guide 1.97.

i The new wide range measurement will use existing containment penetrations.

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The proposed design addition and Technical Specification change does not involve an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-280 SYSTEM: Safety Features Actuation System (SFAS)

COMPONENT: N/A CHANGE, TEST OR EXPERIMENT: This FCR provides a means to perform the monthly functional test of the sequence logic per Technical Specification 4.3.2.1.1.

To perform this test will require the use of spare conductors in existing cables and internal wiring changes in existing equipment.

Work was completed August 24, 1980.

REASON FOR CRANGE:

the previous scheme for testing undervoltage relays and their associated timing relays could end in the tripping of essential 4160 volt buses C1 and DI.

With this modification, an operator is now able to generate a half trip on each essential 4160V.

SAFETY EVALUATION

SUMMARY

This modification has the capability to be easily tested for functional verification and at no time violate the single failure criterion for all protection systems. The implementation of this FCR has had no effect on the nuclear safety of plant equipment or personnel. No unreviewed safety question was involved.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-204 SYSTEM: Containment Purge COMPONENT: Miscellaneous CHANGE, TEST OR EXPERIMENT: This FCR updated drawing E-58B to reflect the following changes:

1)

Changed F259 to read F237 2)

Changed F251 to read F238 This work was completed July 18, 1983.

REASON FOR CHANGE: To correct drawing E-58B and to reflect the conditions in the field.

SAFETY EVALUATION

SUMMARY

This FCR involves correction of design document 7749-E58B Sheet 6 (drawing) in order to correct a drafting error.

The particular change is required in the elementary control diagram of Containment Purge Valve CV-5007 in order to reflect as-built conditions.

This is not an unreviewed safety question.

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a COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-082 SYSTEM: Main Steam i

COMPONENT: PSL-100A&B and PSL-101A&B CHANGE, TEST OR EXPERIMENT: FCR 82-082 allows alteration and replacement of the main steam isolation valve pressure sensing line tubing for pressure switches PSL-100A&B and PSL-101A&B. Work was completed January 10, 1985.

REASON FOR CHANGE: Thermal expansion and movement of the main steam isolation valves (MSIV) caused the tubing to PSL-100A&B and PSL-101A&B to deform. Deformation reduced the integrity of the tubing which is required for pressure sensing of the MSIVs.

SAFETY EVALUATION

SUMMARY

The purpose of the MSIVs is to limit an excessive Reactor Coolant System cooldown rate which could result in reactivity insertion following an air steam line break downstream of the MSIVs. The safety function of the tubing provides a route for the pressure signal required for the cperation of the MSIVs. By replacing the tubing,

.the safety function of the MSIVs will not be degraded, therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-401 SYSTEM: Various COMPONENT: Electrical Penetrations CHANGE, TEST OR EXPERIMENT: FCR 79-401 was requested for the following items:

1)

Engineering review for the price and delivery of replacement modules for the electrical penetration assemblies 2)

Adding a junction box to Electrical Penetration 55 3)

Installing angle adapters on replacement coaxial modules in penetrations PIL1L1 and P2I4G1 4)

Installing Nelson studs in Electrical Penetration P1 CSS 1 and P2CSC1.

Work was completed August 7, 1982.

REASON FOR CHANGE:

1)

To provide engineering evaluation and approval of modification due to engineering evaluation and to purchase spare parts needed i

for modification 2)

To install spare parts due to engineering evaluation SAFETY EVALUATION

SUMMARY

The engineering review and changes to existing penetrations did not reduce system integrity for penetrations modified.

Therefore, an unreviewed safety question does not exist.

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c TOLEDO EDISON March 10, 1986 Log No. KB86-0265 File: RR 2 (P-6-86-02)

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, February 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 February 1986.

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

Yours truly, Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/1jk Enclosures cc:

Mr. James G. Keppler, w/l Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 l

Office of Inspection and Enforcement Mr. Walt Rogers, w/1 NRC Resident Inspector y

LJK/002-I THE TOLEDO E0! SON COMPANY EDISON PLAZA 300 MAGISDN AVENUE TOLEDO, OHIO 43652 l