ML20077H730

Monthly Operating Rept for June 1983

ML20077H730
Person / Time
Site:	Davis Besse
Issue date:	07/08/1983
From:	Caba E, Murray T TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K83-992, NUDOCS 8308110250
Download: ML20077H730 (20)

Text

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OPERATING DATA REPORT DOCKET NO.

50-346 DATE July e, 1983 COMPLETED BY Erdal Caba TELEPHONE 419-259-5000, Ext. 196 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 N0"S

2. Reporting Period:

June, 1983

3. Licensed Thermal Power (MW ):

2772

4. Nameplate Rating (Gross MWe):

925 906

5. Design E!ectrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe):

918

7. Maximum Dependable Capacity (Net MWe):

874

8. If Chantes Oscur 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. If Ar.y:

This Month Yr..to.Date Cumulative iI. Hours in Reporting Period 720.0 4,343.0 43,104.0

12. Number Of Hours Reactor Was Critical 720.0-4.002.8 24.898.3-

13. Reactor Reserve Shutdown Hours 0.0 313.9 3,678.0

14. Hours Generator On Line 720.0 3.950.4 23.710.0

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

= 16. Gross Thermal Energy Generated (MWH) 1.819.447 10,331,316 55,704,077

17. Gross Electrical Energy Generated (MWH) 598,563,,,

3,446,312 18,551,966

18. Net Electrical Energy Generated (MWH) 566.951 3.264.490 17.379.930

19. Unit Service Factor 100_o 90.9 55.0

20. Unit Availability Factor 100.0 90.9 59.0

21. Unit Capacity Factor (Using MDC Net) 90.1 86.0 46.1

22. Unit Capacity Factor (Using DER Net) 86.9 82.9 44.5

23. Unit Forced Outage Rate

0. 0 9.0 18.9

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

Juli 29 1983 Refueling Outage Duration: Approximately 8 weeks

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

26. Units in Test Status iPrior to Commercial Operation):

Forecast Achiesed INITIAL CRITICA LITY INITIA L ELECTRICITY COMMERCIAL OPER ATION 8308110250 830708

("/77 3 PDR ADOCK 05000346 R

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D UNIT SHUTDOWNS AND POW:;t REDUCTIONS Ui NAIE D v s Besse Unit 1 DATE

.Iuly 8 1983 COMPLETED BY Erda1 Caba REPORT MONTH June. 1983 TELErilONE 419-259-5000. Ext. 196 f

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No.

Date k

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jjy Event gy 9-Action to yU Prerent Recurrence-8-

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j;5 Report ar mO 6

I No unit shutdowns or power reduc-tions this month.

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l F: Forced Reason:

Method:

Exhibit G-Instructions S: Schedu!cd A Equipment Failure (Explain) 1-Manual for Preparation of Data 1

B Maintenance of Test 2-Manual Scram,.

Entry Sheets for Licensee C. Refueling

3. Automatic Scram.

Event Repor (LER) File (NUREG-D. Regulatory Restriction

' 4-Cc.itinuation from Previous Month 0161)

E Operator Training & License Examination l 5-Load Reduction F Administrative

9-Other (Explain) 5 f

G-Operational E ror (Explain)

Exliibit 1 - Same Source 19/77)

Il-Other (Explain) 9 I

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OPERATIONAL

SUMMARY

June, 1983 The unit remained at approximately 90 percent reactor power the entire month of June. The power is still limited due to increased condensate flow associated with a realignment of feedwater heater drains following an erosion of moisture separator reheater drain piping to the condenser.

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.v REFUELING INFORMATION DATE: June, 1983 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown: July 29, 1983 3.

Scheduled date for restart following refueling: September 23, 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.

o COMPLETED FACILITY CHANGE REQUEST FCR No: 77-384 SYSTEM: Miscellaneous COMPONENT: N/A CHANGE, TEST OR EXPERIMENT: The specification for field painting, 7749-A-24 was revised October 20, 1978 to allow the use of Amercoat 90, with the thickness specified in Attachment 1 of this specification for Amercoat systems in all areas where nuclear quality paint is required by Attachment 2 of the specification. This includes repair and touch up work, as well as any new items added to these areas.

' REASON FOR CHANGE: Use of one coating system for all areas where nuclear quality painting is required has eliminated the need to stock various types of paint of limited shelf life. Additionally, Amercoat 90 may be applied by brush, roller, or spray as a primer and a top coat over power or hand sanded surfaces.

SAFETY EVALUATION: This change has not adversely affected the safety function of the paint system. A complete engineering evaluation has been made of the flame spread ratings, heat of combustion, and nuclear testing of the Amercoat 90.

The results of this evaluation confirm the acceptabi-lity of this change in the painting specification. No unreviewed safety question was involved.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 77-389 SYSTEM: Process Monitoring COMPONENT: Radiation Elements 2024, 2025, 5029, 5030, 5032, 5403, 5405, 5327, 5328, 1003A, and 1003B CHANGE, TEST OR EXPERIMENT: The enclosure for the above radiation elements and piping was modified so that the bypass and outlet filters for the air pumps are mounted on the outside of the enclosure. A removable cover was fabricated for the filters. Work was completed June 8, 1982.

' REASON FOR CHANGE: Prior to this modification, when the pump was shut

- down for filter cleaning, the unit had to cool for four to six hours as the enclosure becomes very hot. Personnel may now clean the filters in a

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reasonable amount of time without receiving burns.

SAFETY EVALUATION: These changes are an improvement to the system and-will not have an adverse effect on the safety of the plant.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-342 SYSTEM: Pressurizer Spray COMPONENT: Motor operator for valve RC-HV2 CHANGE, TEST OR EXPERIMENT: Velan drawing R35216-4 and any other affected drawings were revised to reflect that pressurizer spray valve RC-HV2 has a 1.6 HP motor operator rather than a 1.0 HP.

The drawing changes were verified May 3, 1983.

REASON FOR CRANGE: Babcock and Wilcox stated in a letter dated June 2, 1978. that records show a 1.0 HP motor should have been supplied. However, it was determined that the use of the 1.6 HP motor may continue with no change as long as no operational problems develop.

SAFETY EVALUATION:_ This drawing change did not create any new cdverse environments by making these as-built revisions and does not constitute an unreviewed safety question.

1 COMPLETED FACILITY CHANGE REQUEST FCR NO:

79-271 SYSTEM: 480 Volt and 240 Volt Motor Control Centers COMPONENT: 480 Volt and 240 Volt Feeders CHANGE, TEST OR EXPERIMENT: On May 3,1982, the work implemented by FCR 79-271 was completed. This involved the replacement of all thermal overload heaters with shorting bars for all Class 1E motor operated valves.

REASON FOR CHANGE: This change made the system more reliable and simpler.

SAFETY EVALUATION: Overload heaters do not have any safety function in our system. Therefore, their exclusion made the system simpler and more reliable. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-137 SYSTEM: Reactor Coolant Syst.en COMPONENT: Reactor Coolant Pumps and Motors CHANGE, TEST OR EXPERIL NT: An interlock was added to the reactor coolant pumps motor start circuit.- This will ensure that the seal return valves, MU59A through D, are open before the reactor coolant pump motor may be started. This was completed June 1, 1982.

REASON FOR CHANCE: A major contributor to reactor coolant pump seal failures had been. operation of the pump with the seal return valve shut.

Although the pump pre-start checklist verifies that seal return flow has been established, there was no assurance that the seal return valve was open_when the motor was started.-

SAFETY EVALUATION: This FCR was designated as nuclear safety related since the FCR package contained a nuclear safety related drawing change notice. The negative pressure boundaries and fire barriers were sealed properly. Hence, this is not an unreviewed safety question.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-168 SYSTEM: Reactor Coolant Pump COMPONENT: Reactor Coolant Pump 1-2-2 Elevation 596' Platform CHANGE, TEST OR EXPERIMENT: Work implemented by FCR 80-158 was completed March 26, 1982. This involved the remov.21 of a 2 inch by 9 inch section of the top flange of a 14WF219 I-beam e-ad a 2 inch by 9 inch section of grating from the Resctor Coolant Pump 1-2-2 596' elevation platform.

REASON FOR CHANGE: This change was necessary because two of the component cooling water return line spool piece flange bolts could not be properly tightened. This was due to the fact that the top flange of the I-beam interferred with the installation of a wrench on these nuts.

SAFETY EVALUATION: Thia FCR work did not degrade the integrity of the platform and did not affect any safety fuctions. It is not an unreviewed safety question.

COMPLETED FACILITY CHANGE REQUEST

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FCR NO: 80-222 SYSTEM: Emergency Diesel Generators COMPONENT: Various CHANGE, TEST OR EXPERIMENT: This FCR was implemented for the installation of the following improved replacement parts for the emergency diesel generators:

New Part Old Part Number Numbe r Description 8470154 8442661 Stubshaft assembly 9515338 8419151 Gear-spring loaded drive 181702 181433

Bolt,

-20 x 1-3/4 This was completed January 4, 1983.

REASON FOR CHANGE: These replacement parts have been recommended by the vendor as a result of the discovery of a 5/8" bolt found to have " backed" out of the spring loaded drive gear.

SAFETY EVALUATION: The changes reflect design product. improvements which should preclude such failures from recurring and thus improve component i

~ reliability. The vendor has supplied a certificate of conformance to certir'y that the components meet original specification requirements. No unreviewed safety question was involved.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-132 SYSTEM: Station Administrative Procedures COMPONENT: Not Applicable CHANCE. TEST OR EXPERIMENT: The following station administrtaive procedures, which were listed in Appendix 13 of the Final Safety Analysis Report, have been deleted:

MP 1401.15 AD 1828.19 HP 1601.02 AD 1830.00 HP 1605.03 AD 1851.00 AD 1828.16 AD 1847.06 This change was verified March 21, 1983.

REASON FOR CHANGE: These procedures were deleted for the following reasons:

1)

MP 1401.05 - Pressurizer Sprey Valve Removal and Replacement This procedure was deleted because it was a compilation of sections of other procedures, all of which are'related to the removal and replacement of safety related valves.

2)

HP 1601.02 - Guides This procedure was deleted as its contents were included in HP 1601.04 and HP 1605.02.

3)

HP 1605.03 - Working Limits for Contamination This procedure was deleted as its contents were included in HP 1601.04.

4)

AD 1828.16 - Inspection Engineering Training 5)

AD 1828.19 - Designated Inspector Training 6)

AD 1830.00 - Inspection Engineering 4

7)

AD 1831.00 - Quality Verification by Station Personnel 8)

AD 1847.06 - Materials Inspection Procedure The last five procedures were deleted as their functions are now met-within the Quality Assurance Division.

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FCR 81-132, Continued Page 2 SAFETY EVALUATION: These changes did not result in the loss of any safety-related function information or instructions. Therefore, the deletions did not constitute an unreviewed safety question, as procedures may be combined or separated to conform with the procedure's plan, see R.G.

1.33-1972, paragraph C.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-034 SYSTEM: Safety Features Actuation System (SFAS)

COMPONENT: Radiation Monitors RE2004, RE2005, RE2006, and RE2007 CHANCE. TEST OR EXPERIMENT: This FCR, which was completed June 28, 1982, called for changes for the SFAS containment radiation monitor trip setpoints and alarm setpoint. The new setpoint for Modes 1, 2, 3, and 4 will be 1.8 times background at the rated thermal power ! 10% of the background reading at the rated thermal power.

In Mode 6, the new setpoints will be 15 mr/hr, 25 mr/hr, 25 mr/hr, and 15 mr/hr for RE2004, RE2005, RE2006, and RE2007, respectively. Guidelines for determination of background radiation at rated thermal power for Modes 1, 2, 3, and 4 have been' incorporated into existing procedure ST 5031.01. Trip setpoints for the monitors in c

Mode 6 were determined by adding 15 mr/hr to the expected background radiation at their repective locations. The 15 mr/hr would be the contribu-tion from a fuel handling accident.

,i REASON FOR CHANGE: This change vill ensure that Technical Specification requirements are not violated. These requirements are two times background j

at the rated thermal power for Modes 1, 2, 3, 4, and 6.

SAFETY EVALUATION: The safety function of the SFAS containment radiation mo:11 tor is to provide the automatic signal for the initiation of containment isolation in the event of a loss of coolant accident. The monitors are located in the annulus during normal operation and moved into containment i

to monitor, in case of a fuel handling accident, in Mode 6.

It has been concluded that this change provides adequate assurance of compliance with Technical Specification requirements.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-044 SYSTEM:

125 Volt DC Essential Power COMPONENT: AC Distribution Panel Y4 4

- CHANGE, TEST OR EXPERIMENT: Resistor RY100, located in essential instrument panel Y4, was temporarily replaced with a resistor with the same electrical characteristics but different physical characteristics. This FCR requested an evaluation of the temporary mounting of the resistor which was completed-June 8, 1982. A permanent resistor was installed July 30, 1982.

REASON FOR CHANGE: The replacement resistor was supplied by Cyberex, Inc., the manufacturer of panel Y4.. However, the original resistor model 4

is now obsolete, making a cutom made permanent replacement necessary to be compatible with the original design.

l SAFETY EVALUATION: Since the electrical values of the replacement resistor were identical with those of the original resistor, the voltmeter character-istics were not affected. The replacement resistor was mounted safely to prevent the creation of an adverse environment in panel Y4.

The ty-rap mounting and the negligible difference between the weights of the original and temporary resistors did not adversely alter the seismic characteristics of the panel. Therefore, the panel retained the capability to perform its i

safety function as required.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-085 SYSTEM: Once Through Steam Generator (OTSG)

COMPONENT: Steam Generator Tubes CHANGE, TEST OR EXPERIMENT: Work implemented by FCR 82-085 was com'pleted August 1, 1982. The purpose was to perform tube plugging and stabilization of no more than ten tubes in each of the OTSGs.

REASON FOR CHANGE: This stabilization helped reassure steam generator tube integrity following the 1982 Refueling Outage.

SAFETY EVALUATION: The safety function of the steam. generator tubes are as follows:

(a) Provides a pressure boundary for the reactor coolant system.

(b) Provides a heat transfer surface for the exchange of heat from the reactor coolant system to the steam generator secondary side.

In regard to item (a), the structural adequacy of this change was verified by Babcock and Wilcox on Sheet 13 of Field Change Authorization 04-3831-00.

This authorization states that with this change, the steam generators meet the pressure boundary requirements of ASME Boiler and Pressure Vessel Code, Section Ill, 1968 Edition, 1968 Addenda, and the original equipment specification.

Item (b), was addressed on Sheet 2 of Facility Change Authorization 04-3831-00 which states that the removal of 300 tubes per steam generator will have no measurable effect on the performance as long as they are randomly distributed.

Prior to this, a total of 14 tubes had been removed from service in Steam Generator 1-2 and 6 from Steam Generator 1-1.

Thus, with the removal of no more than 10 tubes from each steam generator, there is no impact on the performance or reliability of the steam generators.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-103 SYSTEM: Containment Air Sample System COMPONENT: Valves HV5010E and HV5011E.

CHANCE. TEST OR EXPERIMENT: This FCR was implemented to reduce the torque switch settings for Limitorque actuators for valves CV5010E and CV5011E.

the Containment Hydrogen Analyzer #2 and #1, respectively, discharge line valves. The new settings for both opening and closing are 1.0.

Work was completed August 11, 1982.

REASON FOR CHANGE: The original torque switch settings of these two valves caused the valve stems to be overtorqued and, consequently, to bend. The reduced settings will help prevent.this.

SAFETY EVALUATION: The safety function of these valves is to isolate containment on a Safety Features Actuation System (SFAS) Incident Level 1.

The new settings of 1.0, both for closing and opening, have enhanced the equipment operation and are still sufficient for the valves to perform their function. Therefore, this change has not affected the safety function of these valves. Hence, no unreviewed safety question is involved.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-302 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feedwater Pump ClANGE, TEST OR EXPERIMENT: The original speed changer motor addition, by the Terry Turbine Company, consisted of one mounting bracket, a Bodine motor, and a Waldron Type 4 coupling which was 2.25 inches in axial length. A Waldron Type 4 coupling which is 3.5 inches in axial length is now being used. This necessitated the fabrication and installation of a spacer plate between the Woodward Governor Bodine motor baseplate and the motor mounting bracket. Work was completed January 25, 1982.

REASON FOR CHANGE: The Waldron Type 4 coupling, which was 2.25 inches, had cree.ked longitudinally due to insufficient mass to resist the torque applied by the Bodine motor. The use of the longer Waldron Type 4 coupling has eliminated this problem as it contains sufficient mass to withstand the torque of the Bodine motor.

SAFETY EVALUATION: All components of the Auxiliary Feedwater Pumps are Seismic Class 1.

The addition of the longer Waldron coupling and spacing plate has a negligible effect on the seismic analysis. This change hae increased the reliability of the Auxiliary Feedwater Pumps to perform their intended function.

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%m EDISON July 8, 1983 Log No. K83-992 File: RR 2 (P-6-83-06)

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. Italler:

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

Yours truly, M

Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/BMS/ljk Enclosure cc:

Mr. James G. Keppler Regional Administrator, Region III Enc 1:

I copy Mr. Richard DeYoung, Director Office of Inspection and Enforcement Enc 1:

2 copies Mr. Tom Peebles NRC Resident Inspector Enc 1:

1 copy Z~E2y

'fi THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MAOISON AVENUE TOLEOD, OHIO 43652 i

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AVERAGE DAILY UNIT POWER LEVM.,

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DOCKET NO.

50-346 WIT Davis-BessexUnit 1

-f DdTE hly 8, 1983

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COMPLETED BY Erdal Ca TELEPHONE J19-2394000

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N MONTH June, 1983 s

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY PbWER LEVE'.

(MWe Net)

(MWe Net) l 792 37 784 _

wL 793 2

18 787 3

795 g9 705 4

797 20 784 5

796 788

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21 6

798 22 785 7

795 23 785 8

798 24 784 9

788 25 784 10 783 26 784 gg 785 27 784 12 786 782 28 13 787 784 29 786 14 30 782 15 781 3

16 783 INSTRUCTIONS On this format, list the average daily unit power leselin MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77)

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