ML19347B548
| ML19347B548 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/08/1980 |
| From: | Sarsour B TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML19347B545 | List: |
| References | |
| NUDOCS 8010150323 | |
| Download: ML19347B548 (13) | |
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AVERAGE DAILY UNIT POWER LEVEL
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P 50-346 DOCKET NO.
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UNIT Davis-Besse Unit 1 l October 8, 1980 DATE Bilal Sarsour COMPLETED BY (419) 259-5000, TELEPHONE Extension 251 l
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September, 1980 MONTH l
! , DAY AVERAGE DAILY POWER LEVEL ' DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Neti 0 37 0
f 1 0 jg o 2
0 39 0
3 O y 0 4 .
0 21 0 5
0 22 0
6 0 0 7 23 0 24 0 8
0 9 25 0 26 0 10 O 27 o 1I 0 2g 0 12 0 29 0 1 13 (
0 30 0 I4 l 0 3, 15 I l
16 0
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On this format, list the aurage daly unit power lesel in MWe-Net for each day in the reportinginonth. Compute to
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(4/77)
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8010159323
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OPERATING DATA REPORT DOCKET NO. 50-3.46 DATE uccooet _
o, 1980 CONtPLETED BY Bilal Sarsour TELEPHONE (L191 '59-5000,
- OPERATING STATUS r
I. Unit Name: Davis-Besse Unit 1 Notes
- 2. Reporting Period: September, 1980
- 3. Licensed Thermal Power (31Wt): 2772 .
- 4. Nameplate Rating (Gross 3the): 925
- 5. Design Electrical Rating (Net 31We): 906
- 6. Sfaximum Dependable Capacity (Gross 31We): 934
- 7. Sf aximum Dependable Capacity (Net 3the): 890
- 8. If Changes Occur in Capacity Ratings (items Number 3 Through 7) Since Last Report. Gise Reasons:
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- 9. Power Lesel To Which Restricted,if Any (Net 31We)
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- 10. Reasons For Restrictions.lf Any:
This Sionth Yr to Date Cumulative II. Hours In Reporting Period 720 6,575 27,100
- 12. Number Of Hours Reactor Was Critical 0 2,078 13,042
- 13. Reactor Resene Shutdown flours 0 _
0 28,758
- 14. Hours Generator On Line 0 2,008.7 11,883
- 15. Unit Resene Shutdown Hours 0 0 1,728
- 16. Gross Thermai Energy Generated (51WH) 0 4.687.305 24.886.812 .
17 Grcss Electrical Energy Generated ( AthH) 0 1.583.559 8.307.070
- 18. Net Electrica! Energy Generated (>!WH) 0 1.483.787 7.654,365
- 19. Unit Senice Factor 0 30.6 44.5
- 20. Unit Asailability Factor 0 30.6 51.4
- 21. Unit Capacity Factor (Using .\!DC Net) 0 25.4 33.8
- 22. Unit Capacity Factor (Using DER Net) 0 24.9 33.2
- 23. Und Forced Outage Rate 0 14.3 25.6
- 24. Shardowns Scheduled Oser Next 6 Stonths (Type. Date and Duration of Each):
- 25. If Shut Ddwn At End Of Report Period. Estimated Date of Startup: ctober 12, B80
- 26. Units In Test Status (Prior to Commercial Operation): Forecast Achiesed
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INITIA L CRITICA LiTY INITIAL ELECTRICITY j CO\lilERCIA L OPER ATION l l
1 (9/77)
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50-346 DOCKET NO. .
UNIT S!!UTDOWNS AND I OWE!! REDUCTIONS -
UNIT NAME Davia-B+'sne Ifnit 1 October 8, 1980
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DATE COMPLETED BY Bilal Sarsour REPORT MONTli
- september, 1980 TELEPl!ONE (419) 259-5000, Extett-sion 251
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-, .$ ? 3 $ .E 5 Licensee P, u'g {*3
, Cause & Corrective Action to No. Date g 3g s .3 s s Event 9 H g $ Report a mu 8O Prevent Recurrence 3i s @ ; =E u 6
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8 4 NA NA Nlt The unit outage which began on April 4 30 04 7 S 720 C
- 7, 1980, was still in progress through the end of September, 1980.
See Operational Suminary for further details.
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! 2 3 4 F: Forced Reason: Method: Exhibit G. Instructions A. Equipment Failure (Explain) I Manual for Preparation of Data S: Schedu!cd B. Maintenance of Test 2 Manual Scram. I:ntry Sheets for Licensee C Refueling 3 Automatic Scram. Event Report (LER) Fife (NUREG.
D.ReEutatory Restriction 4.Other (Explain) 0161)
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!?Operatos Training & Licetisc Examination 12-Adunimt rative 5
- G 0;vr.aional Enror (Explain) Exlubit I - Same Source
- (9/77) II.Other (lisplain)
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OPERATIONAL SLMfARY SEPTEMBER, 1980 The unit outage which began on April 7,1980, was still in progress through the end of September,1980 with hanger modifications required to satisfy NRC Bulle-tin 79-02 and 79-14 being the critical work.
In addition sevet<_ snubber mountings required modifications.
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The Reactor Coolant System is full and vented and a bubble was drawn in the pressurizer. Hanger work and diesel generator work prevented system heatup.
During preventive maintenance on the emergency diesel generators, a bolt fragment was found in Diesel Generator 1. The bolt fragment came from the back of the gear in the camshaft / turbocharger gear train. A new gear with positive locking was installed on both diesel generators.
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REFUELING INFO?JtATTC'i DATE: September, 1980 i
- 1. Name of facility: Davis-Besse Nuclear Power Station Unit 1 March, 1982
- 2. Scheduled date for next refueling shutdown:
May, 1982
- 3. Scheduled date for restart following refueling:
- 4. Will refueling or resumption of operation thereaf ter 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
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to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?
Reload analysis not completed, none identified to date.
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- 5. Scheduled date(s) for submitting proposed licensing action and supporting information. January, 1981
- 6. Important licensing considerations associated with refueling, e.g., new or different fuci design or supplier, unreviewed design or performance analysis methods, significant changes in fuci design, new operating procedures.
I None identified to date.
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- 7. The number of fuel assemblics (a) in the core and (b) in the spent fuel storage pool.
44 - Spent Fuel Assemblies (a) 177 (b) . 8 - New 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 pool assuming the present licensed capacity.
Date 1988 (assumins ability to unload the entire core into the spent fuel
pool is maintained)
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COMPLETED FACILITY CHANGE REQUESTS ;
i FCR NO: 77-418 i SYSTEM: Main Steam Supply COMPONENT: MS1Vs
. CHANGE, TEST, OR EXPERIMENT: On March 19, 1980 all inner lens on the Cutler-Hammer E30DX1 pushbutton units were replaced by a special lens which acts as an insulator between the lamp and the metal frame
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REASON FOR THE CHANGE: When removing the lamp, a short circuit between the lamp terminals and the frame assembly may occur causing the fuse to blow.
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SAFETY EVALUATION: The modification covered under this FCR involves the replacement of the inner lens on E30DX1 Cutler-Ilammer pushbutton units with a special lens (Catalog No. E30YED129). ?bny Q circuits employ this pushbutton. This modification does not require any circuits to be de-energized to accomplish the work. Installation of the new insulating inner lens will not create any new adverse environment and will enhance the reliability of the circuits by eliminating cause for blown fuses when bulbs are replaced. No unreviewed safety question exists.
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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-091 SYSTEM: Diesel Generators 1-1 and 1-2 COMPONENT: Reverse Power Relay 32/DG CHANCE, TEST, OR EXPERIMENT: On May 8, 1980' work was completed on FCR 79-091 which revised the relay pickup and time dial settings for the reverse power relay used in the Diesel Generator protection scheme. The relays on both Diesel Generators 1-1 and 1-2 will now operate at the minimum pickup of 0.3 amperes, and the time dial was
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readjusted to number 4 setting which corresponds to an 11 second time delay.
REASON FOR THE CHANGE: These modifications will ensure that there is ample time to synchronize the diesel generator and apply load to the bus before the reverse power relay picks up.
SAFETY EVALUATION: This FCR consists of changing the relay pickup and time dial setting for the reverse power relay used in the Diesel Generator protection scheme.
This change will not affect the safety function of the Emergency Diesel Generator.
It will improve the time required to synchronize the diesel generator and apply load. This change is internal to the relay and therefore no adverse environment is created. This is, therefore, not an unreviewed safety question.
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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-139 SYSTEM: Main Steam COMPONENT: Main Steam Safety Valves SP17A1, SP17A2, SP17A8, SPl7A9, SP17B1, SP17B2, SP17B8, and SP17B9
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CHANCE, TEST, OR EXPERDIENT: On April 10, 1979 the position change of the 1050 psig valves with the 1100 psig valves found on the east and west headers of the main steam line was completed. The position of valve SP17Al was switched with SPl7A8; SP17A2 with
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SP17A9; SP17B1 with SPl7B8; and SP17B2 with SP17B9.
REASON FOR THE CHANGE: There have been repetitive problems with the low range (1050 psig) safety valves. Responsible personnel from Babcock and Wilcox suspect that flow vibration or flow characteristics may cause the valve to lif t early since they are located close to a bend in the pipe.
SAFETY EVALUATION: The requested change to relocate four of the code safety valves on each of the main steam headers will make no change to the relieving capacity of the valves so moved or the total relieving capacity of the header. The setpoints of the valves, nor any of the other features of the valves are to be changed. The method of venting the valves to atmosphere will not be changed.
This is not an unreviewed safety question.
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- l COMPLETED FACILITY CHANGE REQUESTS 1
i i FCR NO: 79-355
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f SYSTEM: 480 Volt Power Supply System
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COMPONENT: Power Operated Relief Valve (PORV) RCll 1
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} CHANGE, TEST, OR EXPERIMENT: On December 14, 1979 work was completed on FCR 79-355.
l This FCR provided the capability for the PORV RCll to bc supplied with control and
- motive power from the energency buses. This was accomplished by (1) upgrading the i control and native power source to Class IE and (2) upgrading of the installation j inside containment to Class IE.
In addition, the circuit breaker numbers and associated control scheme for the 480 j volt power supply were revised and procedures SP 1103.05 and SP 1107.07 were modified j to incorporate these changes.
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- REASON FOR THE CHANGE
- These changes were made to comply with the TMI-2 Lessons j Learned Task Force Report (short term), NUREG-0578.
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j SAFETY EVALUATION: The function of the PORV block valve RCll is to prevent the 1 depressurization of the reactor coolant system by isolating the electromatic relief
- valve if it has stuck open.
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By upgrading the power supply to the block valve operator to Class IE, the normal
- function of the block valve is not affected. This change will facilitate the opera-i tion of the block valve during a loss of offsite power since the valve operator will j then be supplied from an essential bus.
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- This change will not affect the Class IE source (MCCF12A) as the power circuit for
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this valve can be separated from the MCCF12A by a Class IE breaker. The 120 volt side of the control circuit is isolated from the 480 volt power side by the control transformer. Ground fault protection exists on the neutral of the transformer in j the unit substation F1 which supplies power to MCCF12A.
} An unreviewed safety question does not exist.
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I COMPLETED FACILITY CHANGE REQUESTS i
L 1 FCR No: 79-429 SYSTEM: Instrument Ground System
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j COMPONENT: N/A
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CHANGE. TEST, OR EXPERIMENT: On June 26, 1980 work was completed on FCR 79 *.29.
This FCR provided (1) testing of the instrument ground system for the detection of any jnadvertent connections with the station ground system and (2) documentation of any inadvertent connections between the instrument and station ground systems for further analysis and correction.
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REASON FOR THE CHANCE: Results of testing the instrument ground system had shown 4
that there could have been inadvertent ties between the instrument and station ground i systems at other than the designated common tie point.
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l l SAFETY EVALUATION: This FCR provides for the testing of the instrument ground system for the detection of inadvertent connections with the station ground system. Inad-
] vertent connections will be documented for further analysis and correction. This testing will help in meeting Toledo Edison's commitments to the NRC. This FCR is nuclear safety related because it requires testing to be performed on Class IE equip-ment. This testing will not create a new adverse environment and does not constitute j an unreviewed safety question. t i !
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COMPLETED FACILITY CHANGE REQUESTS FCR No: 79-445 SYSTEM: High Pressure Injection System
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COMPOSENT: Hanger 33C-CCB-2-H46 CHANGE, TEST, OR EXPERIMENT: On December 30, 1979 work was completed on FCR 79-445 which was comprised of the modification of hanger 33C-CCB-2-H46. This hanger is located on the 2 " discharge line from the high pressure injection pump.
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REASON FOR THE CHANCE: Refined analysis of this hanger indicated that the slender-ness ratio was less than the design criteria.
SAFETY EVALUATION: The modifications to this hanger will assure that the design criteria slenderness ratio of 200 is satisfied. This will assure hanger performance in accordance with design criteria. There will be no adverse effect on the high pressure injection system performance. Therafore, this is not an unreviewed safety ques t ion.
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4 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 80-030 SYSTEM: Reactor Protection System / Nuclear Instrumentation COMPONENT: NIM-BINS CHANGE, TEST, OR EXPERIMENT: On April 18, 1980 work was completed on FCR 80-030 which provided the temporary installation of the NIMBINS NI-1 and NI-2, source range nuclear instrumentation. The NIMBINS provides an audible indication of the neutron
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count rate from containment to the control room.
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REASON FOR THE CHANGE: Prior to entering Mode 6, the station must comply with Tech-nical Specificatica 3.9.2, Refueling Operations. This technical specification states that two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in containment and the control room.
SAFETY EVALUATION: This FCR provides a temporary connection from a reactor protection system cabinet to a counter scaler for the purpose of providing an audible indication of increasing neutron icyc1 in containment during refueling.
This is nuclear safety related as the temporary cabic will be routed to a nuclear safety related cabinet.
Since the installation only exists at a time when the reactor is tripped, the temporary cable will not af fect the operation of u.- nactor protection system as it is not required in Modes 5 or 6.
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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 80-123 SYSTEM: Makeup and Purification System COMPONENT: MU242, MU243, MU244, and MU245 CHANCE, TEST, OR EXPERIMENT: On June 7, 1980 modifications on valves MU242, Mti243, d MU244 and MU245 for FCR 80-123 were completed. The modifications made to tra valves were grinding out the canopy and seal welding the bonnet to the valve using a fillet
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REASON FOR THE CHANGE: In order to remove the bonnet for valve repair, the canopy weld must be ground out. Tolerances for rewelding the canopy cannot be met to allow f reassembly via a canopy weld.
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SAFETY EVALUATION: This FCR involves changing drawings to allow installation of a valve with a fillet type seal weld for valves MU242, MU243, MU244, and MU245. Seal-ing the bonnet to the body with a fillet weld does not affect the ability of the valve to perform its safety function. This valve modification has the concurrence
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of the valve manufacturer, Rockwell International. An unreviewed safety question does not exist.
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