ML20154S568

Monthly Operating Rept for Dec 1985

ML20154S568
Person / Time
Site:	Davis Besse
Issue date:	12/31/1985
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
KB86-0012, KB86-12, NUDOCS 8604010323
Download: ML20154S568 (18)

Text

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7 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit 1 DATE January 10, 1986 COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000 Ext. 7290 MONTH December 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) IMWe-Neti I O -

37 0 2 0 0 18 3 0 19 0 4 0 20 0 5 0 21 0 6 0 22 0 7 0 23 0 8 0 24 0 9 0 23 0 10 0 26 0 II O 27 0 12 0 28 0 13 0 0 29 14 0 30 0 15 0 3g o 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) 8604010323 851231 PDR R

ADOCK 05000346 PDR

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OPERATING DATA REPORT DOCKET NO. ' 50-346 DATE Janunty 10, 1986 COMPLETED BY Mortezn Khazrai TELEPHONE (419) 249-5000 OPERATING STATUS Ext. 7290

1. Unit Name: Davis-Besse Unit 1 Notes

. 2. Reporting Period: December 1985

3. Licensed Thermal Power (MWr): 2772

4. Namepizte Rating (Gross MWe)- 925

5. Design Electrical Rating (Net MWeI: 906

6. Maximum Dependable Capacity (Gross MWe): 904

7. Maximum Dependable Capacity (Net MWe): 860

8. If Char.3es Occur in Capacity Ratings (items Number 3 Through 7) Since last Report. Give Reasons:

9. Power Level To Whici. Restricted,if Any (Net MWe):

10. Reasons For Restrictions.If Any:

This Month Yr to.Date Cumulative

11. Hours in Reporting Period 744.0 8,760.0 65,065

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 (SIWH) 0.0 6.312.I78 81.297.600 .

17. Gross Electrical Energy Generated (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 31.2 52.8

20. Unit Asailability Factor 0.0 31.2 55.5

21. Unit Capacity Factor iUsing MDC Net) 0.0 25.8 45.I

22. Unit Capacity Factor IUsing DER Net) 0.0 24.5 42.8

23. Unit Forced Outage Rate 100.0 64.8 25.3

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:

26. Units in Test Status (Prior o Commercial Operationi: Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIA L OPER ATION N/77I

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DOCKET NO. 50-346 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit 1

• DATE January 10. 1986 .

COMPLETED BY Morteza Khazrai REPORT MONTH December 1985 TELEPHONE (419) 249-5000. Ext. 7290 L

"u jg } $ Licensee a 4, g Cause & Corrective No. Date g gg g yuy Event g7 gg Action to s uo e u :s Report # xo au Prevent Recurrence 85 &#9

• 8 8

7 85 06 09 F 744 A 4 LER 85-013 JK SC The unit remained shutdown follow-Con't ing the reactor trip on June 9, 1985.

See Operational Summe.cy fo.- further details, i

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1 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 1 - Same Source I

(9/77) H-Other (Explain) i

OPERATIONAL-

SUMMARY

DECEMBER 1985 __

The unit remained shutdown the entire month of December following the reactor trip on June 9, 1985. Investigation of the causes of the event and corrective actions continue. See NUEEG 1154 for further details.

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

1) Continued MOVATS testing.

2) All work completed on Decay Heat Loop #2.

3) Power operated relief valve was removed, modified, and replaced.

4) Steam generator manway studs were replaced with a design which permits tensioning rather than torquing of the bolts.

5) Motor driven feed pump was installed and testing has begun.

REFUELING INFORMATION DATE: December 1985

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

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, 1986

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

new or different fuel design or supplier, unreviewed design or performance analysin 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 fuci 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 tool assuming the present licensed capacity.

Date: 1992 - assuming ability to unload the entire core into the Spent fuel pool is maintained.

BMS/005

COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-263 ,

SYSTEM: Reactor Coolant System COMPONENT: FI-4102, FI-4202, FI-4302, FI-4402 CHANGE' TEST OR EXPERIMENT: This FCR installed restraints on the lower support piping for the four Reactor Coolant System (RCS) flow indicators listed above. Work was completed. February 21, 1981.-

REASON FOR CHANCE: These flow indicators were exposed to stress due to the lack of-restraint on the lower support piping. This has caused several of the~RCS flow indicators to leak.

SAFETY EVALUATION SEKMARY: The safety function of these flow indicators is to monitor RCS flow which is a parameter that must be monitored closely.

The above change will prevent the leakage of the flow indicators-and allow for a more accurate RCS flow indication. Therefore, an unreviewed safety question does not exist.

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• COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-319 SYSTEM: Main Steam -

COMPONENT: MS-107, CDF11A-2, and CDF11D CHANGE, TEST OR EXPERIMENT: FCR 78-319 initiated two changes. The first change was made to updatet associated drawings involving the circuitry for MS-107 in CEF11A-2 which was necessary to clear jumper and lif ted wire tag 1175. The second change made by FCR 78-319 was to verify the as-built conditions for disconnect switches CDF11A-2 and CDF11D. Work was completed March 30, 1985.

REASON FOR CHANGE: The above changes were made to update the drawings associated with the disconnect switches to represent the correct configura-tioa of the plant.

SAFETY EVALUATION

SUMMARY

The safety function of the disconnect switch cabinet is to provide local control of equipment in case of losing control capability from the Control Room. The change made by this FCR does not effect this safety function. Therefore, an unreviewed safety question' s

does not exi-t.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-016. Rev. A SYSTEM: Auxiliary Building Fire Detection System, Elevation 585' I

COMPONENT: j CHANGE, TEST OR EXPERIMENT: FCR 79-016 Revision A modifies the Fire Detection System on Elevation 585'.

Fire detectors were installed in Rooms 300, 304, (310-313), 312, 321, 322, 325, 328, 303, 314, the annulus space in the area of the electrical i penetrations. Panel C3630 in Room 324, and fira detectors were relocated ,

in Room 323. The modifications to'the Fire Detection System were started November 6,1979, and'the job task was completed January 1,1980.

REASON FOR CRANGE: Detectors were added to Room 301 which adjoins Room .

300 to ensure early detection of fires. Detectors were added to Rooms 323 and 324 due to pockets created by structural steel. One detector was i added to Room 320 due to a wall being added. Detectors were added to i Rooms 318 and 319 to provide fire protection while modifications were j being implemented. One detector was rdded to Room 320A and 321A. Five  ;

detectors were added to the Intake Structure. This includes Rooms 51, 52, and 53.

4 These changes were made to satisfy commitments addressed ir. the Fire Hazard Analysis Report.

SAFETY EVALUATION

SUMMARY

All work was installed with the "Q" core drill

. reports. Post Inspection Construction Authorizations (PICA) insure those portions from creating any new adverse environments. An unreviewed safety question is not. involved.

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U CCMPLETED FACILITY CHANGE REQUEST FCR NO: 80-079 SYSTEM: Main Steam COMPONENT: Various Hangers CHANGE, TEST OR EXPERIMENT: This FCR was implemented to modify the bargers listed below. These hangers are support for the Main Steam System. Work was completed February 2,1984. ,

3A-EBB-2-H9 3A-EBD-19-H100 3A-EED 19-H129 3A-EBB-2-H21 3A-EBD-19-H101 3A-EBD-19-H143 3A-EBB-2-H37 3A-EBD-19-H104 3A-EBD-19-H146 3A-EED-19-H10 3 A-EBD-19 -H105 3A-EBD-19-H151 3A-EBD-19-H47 3A-EBD-19-H106 3A-EBD-20-H2 3A-EBD-19-H48 3A-EBD-19-H107 SR1 East & West 3A-EBD-19-H58 3A-EBD-19-H108 SR2 East & West 3A-EBD-19-H65 3A-EBD-19-H109 SR4 East & West 3A-EBD-19-H67 3A-EBD-19-H112 SR5 Eact & West 3A-EBD-19-H85 3A-EBD-19-H113 SR6 East & West 3A-EBD-19-H92 3A-EBD-19-H115 SR7 East & West SR8 East & West SR18 East & West SR19 East & West SR47 East & West H3 East & West REASON FOR CHANGE: Modifications were made in accordar.ce with IE Bulletins 79-02 and/or 79-14.

SAFETY EVALUATION

SUMMARY

The function of the listed hangers is to support the Main Steam System. Modifications were made to the hangers to reduce stress levels to acceptable values, thus increasing the margin of safety. Therefore, an unreviewed safety question does not exist.

COMPLETED FACILITY CHANGE REQUEST FCR N0: 81-058 SYSTEM: Emergency Diesel Generators COMPONENT: K5-1 and KS-2 CHANGE, TEST OR EXPERIMENT: Thin FCR was implemented to modify the engine control system of the Emergency Diesel Generators (EDGs) to utilize the engine speed control motor (governor motor) on the mechanical gcvernor to regulate fast start acceleration. This modification changed the 0-900 rpm time from 4-5 seconds to the vendor's recommendation of 8-9 seconds.

Technical Specifications require EDGs to start from ambient condition and accelerate to at least 900 rpm in less than' or equal to 10 seconds. Work was completed August 30, 1983.

REASON FOR CHANGE: This modification will prolong the life and reliability of the turbocharger and gear train of the EDGs.

SAFETY EVALUATION

SUMMARY

The safety function of the EDGs is to provide on-site standby power sources for essential loads required for safe plant shutdown. By increasing the response time of the EDGs, their safety function will be maintained and their reliability will increase. This will not create an adverse environnent. Therefore, this does not constitute an unreviewed safety question.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-050 SYSTEM: Reactor Protection System COMPONENT: Cable 4LRPSC07 CHANGE, TEST OR EXPERIMENT: This FCR was initiated for the retermination of cable 4LRPSC07 on the leads of Penetration Module C, and to test the cabic for resistance of > 10p. Work was completed June 18, 1982.

~ REASON FOR CHANGE: During field checks of cable resistance to ground penetration P4LIG, low resistance was encountered which caused FCR 82-050 to be initiated.

SAFETY EVALUATION

SUMMARY

The safety function of cable 4LRPSC07 is for reactor coolant loop and hot leg narrow range temperature for reactor protection on reactor coolant high temperature. This change has not degraded the safety function of the cable because it was reterminated to a Class IE module and was routed to maintain proper safety channel separation and channel designation. There is no unreviewed safety question.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 83-102 SYSTEM: Reactor Coolant System COMPONENT: Reactor Coolant Pump Piping CHANGE, TEST OR EXPERIMENT. FCR 83-102 was initiated to modify various supports used to maintain the Reactor Coolant Pumps (RCP). Modifications included:

1) deleting support M-1089/H3 on RCP 1-2-1 standpipe flush line from all associated Station drawings

2) modifying the adjacent supports on piping located on RCP 1-2-1

3) removing support on the 3/4" CCB-7 line located on RCP 1-1-1 Work was completed December 19, 1984.

REASON FOR CHANGE: During an inspection, it was discovered that support M-1089/H3 was never installed as designed and, due to obstructions, cannot be installed. For this reaaon, support M-1089/H3 was deleted from the RCS drawings and adjacent supports on the piping had to be modified. Also, a walkdown required by IEB 79-14 identified interferences (during operation) with an RCP 1-1-1 wire rope whip restraint and the 3/4" CCB-7 line located on RCP 1-1-1. This requires the removal of the support on the 3/4" CCB-7 line.

SAFETY EVALUATION

SUMMARY

The safety function of the above listed supports is to maintain the integrity of the RCPs. The above modifications will allow this function to be performed. Therefore, an unreviewed safety qucstion does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-065 l

SYSTEM: Reactor Coolant System l COMPONENT: RC-49 CHANGE, TCST OR EXPERIMENT: 'his FCR allowed for the performance of Section 7.4 of Test Procedurr. . 600.13, the. Pressurizer Operational and Spray Flow Test. This was dot' to reset the flow for RC-49, pressurizer mini-flow valve, as recommended by Babock & Wilcox. The recommended setting range was between 0.75 gpm to 3.0 gpm. Work was completed January 20, 1985.

REASON FOR CHANGE: In the process of connecting and disconnecting the electrical cables to the pressurizer heater bundles, many of the individual heater electrical pins and ceramic insulators were damaged. Damaged heater connectors render the heaters inoperable. Until the connectors are repaired, the required heating capacity during steady state conditions should be reduced, if possible.

The pressurizer's silicon controlled rectifier (SCR) heater bank was sized to provide enough heating capacity to compensate for normal heat loss (ambient losses) and spray valve bypass flow during steady state operations. The original plant design did not require any additional heaters except during load changes and reactor startup.

Presently, the SCR heater bank cannot supply the required heating capacity during steady state operation. Additional heater banks are energized which leaves less capacity available for transient conditions. It is suspected that RC-49 was set above the 0.75 gpm to 3 gpm flow recommended

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by Babcock & Wilcox, thuc, caur,i'g an additional hetter requirement during normal operating conditions.

SAFETY EVALUATION

SUMMARY

_The safety function of the mini-flow valve RC-49 is to eliminate the abnormal buildup or dilution of boric acid within the pressurizer and to minimize cooldown of the coolant in the spray and surge lines. The resetting valve RC-49 will not prevent it from performing its intended safety function. Therefore, an unreviewed safety question does not exist.

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COMPLETED FLCILITY CHANGE REQUEST FCR NO: 84-219 SYSTEM: Reactor Coolant System COMPONEN[: Piping support PSU-R1/PSU-H1 CHANGE, TEST OR EXPERIMENT: FCR 84-219 modified piping support PSU-R1//PSU-H1 located on the 10" pressurizer surge line. Modifications included:

1-) the replacement of two 1 " bore hydraulic snubbers with one 2 "

bore hydraulic snubber,

2) the installment of a kicker brace under the existing built up beam of PSU-H1,

3) the removal of the upper, west section of the structutal tubing from PSU-R1, and

4) the revision o'f the hot / cold settings of 'the spring hanger located on PSU-Hl.

k'ork 'was completed January 3,1985.

REASON FOR CHANCE: Investigation showed tle redesign of the piping support was needed to compensate for the thermal movements of the piping.

SAFETY EVALUATION

SUMMARY

The safety function of the piping support on the 10" surge line is to insure the integrity of the surge piping to the presserizer and, therefore, maintain the RCS pressure boundary under sustained, thermal, and seismic loading. Originally, the thermal movements for this line were not considered in the line's design. The modifications listed above account for the thermal movement of the line, therefore, insuring surge piping integrity. The modifications made do not cause an unreviewed safety question.

COMPIITED FACILITY CHANGE REQUEST FCR NO: FCR 84-221 SYSTEM: Fuel Handling Area Exhaust COMPONENT: Support 410-03-7 CHANGE, TEST OR EXPERIMENT: FCR 84-221 modified the Fuel Handling Area Exhaust System support 410-03-7. Modifications included the removal of a tubing section in column RA-7-4 and the addition of stiffener plates to the underside of the 1/8" plate welded to the bottom of the duct. Work was completed December 31, 1984.

REASON FOR CHANGE: The above modifications were made to increase the duct / support flexibility insuring compliance with design requirements for plant operation.

SAFETY EVALUATION

SUMMARY

The safety function of support 410-03-7 is to provide a supporting / stabilizing function to the Fuel Handling Area Exhaust System during normal operation and a seismic event. By performing this function, the structural integrity of the ductwork is maintained.

The above modifications will enhance this function. Therefore, an unreviewed safety _ question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-224 SYSTEM: Emergency Ventilation COMPONENT: Support 410-07-C-31Z CHANGE, TEST OR EXPERIMENT: FCR 84-224 was implemented to modify Emergency Ventilation System (EVS) support 410-07-6-31Z. The modification involved cutting and removing a 2" x 2" x 1/4" structural angle.from the support.

The vork was completed December 18, 1984.

REASON FOR CHANGE: A'fter system reanalysis, it was determined that the subject support required modification to comply with short and long term operability requirements.

SAFETY EVALUATION

SUMMARY

The purpose of the EVS is to insure that a negative pressure exists in the electrical and mechanical penetration rooms and the containment annulus, and to remove possibly contaminated air from these areas and discharge the air to the atmosphere through the station vent. The safety function of support 410-07-C-312 is co provide a supporting / stabilizing function to the ductwork during both normal operation and a postulated seismic event.. By performing the above. modification, the flexibility and structural integrity of the duct and support will be enhanced which will insure compliance with short and long term design criteria for plant operations. Therefore, an unreviewed safety question does not exist.

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i COMPLETED FACILITY CHANGE REQUEST FCR NO: 84-225~

SYSTEM: Containment Purge Exhaust COMPONENT: 435-04-1B CHANGE, TEST OR EXPERIMENT: This FCR modified the heating, ventilation, and air conditioning (HVAC) ductwork support 435-04-1B located on the "Q" portion of the Containment Purge Exhaust. System. Modifications included adding two extension plates to each upper _ support plate of support 435-04-1B and installing horizontal bracing from the support frame to the Shield Building'. Work was completed December 27, 1984.

REASON FOR CHANGE: The above modifications were made to stabilize the support, thus, insuring compliance withLdesign requirements for plant operation.

SAFETY EVALUATION

SUMMARY

The safety function of of the containment  ;

purge exhaust support 435-04-1B.is to provide restraining / stabilizing action to the ductwork during a postulated seismic event. By performing this function, the structural integrity,of the ductwork and the isolation of the negative pressure boundary are maintained. The modifications made by this FCR will enhance the safety function of the support. Therefore, an unreviewed safety question does not exist.

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TOLEDD w EDISON January 10, 1986 Log No. KB86-0012 File: RR 2 (P-6-85-12)

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

Dear Mr. Haller:

Monthly Operating Report, December 1985 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 December 1985.

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

Yours truly, t

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Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/ljk Enclosures cc: Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Walt Rogers, w/l NRC Resident Inspector n \

LJK/002 THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652 l