ML19350D234

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Monthly Operating Rept for Mar 1981
ML19350D234
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 04/08/1981
From: Sarsour B
TOLEDO EDISON CO.
To:
Shared Package
ML19350D232 List:
References
NUDOCS 8104140379
Download: ML19350D234 (11)
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Text

{{#Wiki_filter:* AVERAGE DAILY UNIT POWER LEVEL O . 50-146 DOCKET N,0. 11 NIT Davis-liesse Unit 1 D A'I E April 8, 1981 CON!PLETED BY ._Bilal Sarsour TELEPl!ONE (419) 259-5000, Extension 251 March, 1981 MONTil DAY AVER AGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (SIWE-Net) (51We Net) 

       ,                        883                                                     659 37 2                        856                                18                   667 3                        547                                39                   661 4                        870                                20                   660 5                        886                                21                   692 6                        840                                22                   807 7                        877                                23                   806 g                         88                                24                   810 9                          0                                                     808 25 10                           0                                26                   807 11 0                                27                   809 12 0                                                     806 2S 13 73 29                . 808 g4 410                                                     805 30 15 474                                3,                   804 16                         608 INSTRUClIONS On this forinat list the average daily unit pimer leselin MWe Net for caeh day in the reporting inonth. Compute to the nearest whole mepwatt.

(9/77i e

O OPERATING DATA REPORT DOCKET NO. 50- M DATE April 8, 1981 COilPLETED !!Y 1;ilal Sarsour TELEPliONE (419) 259-5000, Extension 251 OPERATING STATUS Davis-13 esse Unit 1 Notes

1. Unit Name:
2. Reporting Period: Ma rc h . 1981
3. Licensed Thermal Power t.\lWt): 2772
4. Nameplate Rating (Gross 31We): 925
5. Design Electrical Rating (Net 31We): 906
6. h!aximum Dependable Capacity (Gross SIWE): 934
7. f.laximum Dependahle Capacity (Net Slwe): R90
8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7)Since Last Report.Gise Reasons:
9. Power Lesel To Which Restricted. If Any (Net alwe):
10. Reasons For Restrictions,if Any:

This alonth Yr..to-Dat e Cumu?atise i1. Ilours la Reporting Period 744 2,160 31,469

12. Number Of Ilours Reactor was Critical 709.2 1,462.2 15,846.4
13. lleactor Resene Shutdown flours 34.8 34.8 2,916.9
14. Ilours Generator On-Line 616.6 1.352.6 14.400.4
15. Unit Resene Shutdown liours 0 0 1,731.4
16. Gross 'l hermal Energy Generated.(31Wil) 1,437,670 3,234,198 30,139,004
17. Gross Electrical Ener/ y Gener,ted (.\lWil) 478,048 1,080,766 10,056,100
18. Net Electrical Energy GenerateJ (5thil) 446,468 1,010,370 9,274,871
19. Unit Service Factor 82.9 62.6 46.5
10. Unit Availability Factor 82.9 62.6 52.3
21. Unit Capacity factor IUsing AIDC Net) 67.4 52.6 35.0
22. Unit Capcity Factor IUsing DER Net) 66.2 51.6 34.4
23. Unit Forced Outage Itate 17.1 37.3 26.0
24. Shutdowns Scheduled Oser Next 6 51onths (Type. Date.and Duration of Eacht:
25. If Shut Down At End Of Report Period. Estimated Date of Startup:
26. Units in Test Status IPrior to Commercial Operation): For ecast Acl.iesed INITI \ 1. CRI TICA LITY INITI A L El.l:Cl RICITY C051\lERCI A L OPE R ATION P00R ORIGINAL il 4

9 50-346 i DOCKET NO. _ UNIT S!!UTDOWNS AND POW:.:t REDUCTIONS - UNIT N AME D3Vis-be^Se U"it 1 - DATE Anril R. 198L  ! i COMPLETED IIY Bital sarseur

  • REPORT MONTIl March. 1981 TELEPi!ONE (419) 259-500n. Ext. 251
  • 1 e.

_ E E jg Cause & Coirective o. 3 4.E 5 Licensee Event H t, u7 1, 

                                                                                                              $1 e-                        Action to W.            Date               i2          @   .s ~ $

N'0 C j@ $ jgg Report n yV P.cvent Recurrence 6 NA NA NA NA Power was reduced to approximately 2 81 03 03 S 0 B 55% to replace a single twisted pair cable on Main Feed Pump Turbine #1. - 3 81 03 08 S 0 B NA NA NA NA Power was reduced to 8% to investi-gate the hip.h combustible gas alarm

  • received fro:n the main transformer.

NA NA NA The reactor was manually tripped 4 81 03 11 F 127.4 A 1 following a full Steam and Feed-'ater . Rupture Control System trip. See Operational Summary for further details. T C3 Q .

t2 .

C' Q . 3 4 2 Exhibit G - Instructions Method: g F: Forced S: Schecuted p Reason: A Eituipment Failure (Explain) 1-Manual for Preparation of Data 

          %                                                                                         2-Manual Scram,                   entry Sheets for Liecnsee B. Maintenance oi Test                                                                Event Report (LER) File (NUREG-3-Automatic Scram.

C.Refueline +Ghef-{ Esp'eshH 0161) M D-Regulatdry Restriction E-Operator Training & License Examination 4-Continuation 

     ,====                                                                                                                        5 F- Ad minht rat ive                                 5-Peduetion                                 ~'"*'      "#

G-0;>erational Eiror (Explain) 6-Other 11 Other (E splain) (9/77)

OPERATIONAL

SUMMARY

MARCil, 1981 3/1/81 - 3/3/81 Reactor power was maintained at between 99% and 100% of full power with the turbine generator gross loan at approximately 925 10 MWe. Pouer was reduced to 55% at 0212 hours on March 3, 1981 to replace a single twisted pair cable (LIP) on Main Feed Pump Turbine 01FPT) #1. The reactor power level was maintained at 55% with the genera-tor gross load at approximately 500 t 10 MWe until 0520 hours on March 3, 1981 when reactor power was increased to 99% full . power. 3/4/81 - 3/11/81 The reactor power was maintained at 99% full power with the generntor gross load at 921 1 10 MWe until 0645 hours on March 

                      '8, 1981 when the turbine generator was taken off line to inves-tigate the high combustible gas alarm received from the main transformer, but the reactor stayed critical at approximately 8% power.

The reactor power was maintained at 8% full power until 0845 hours on March 11, 1981 when the reactor was manually tripped following a full steam and feedwater rupturc control system (SFRCS) actuation on low OTSG level. 3/12/81 The reactor was critical at 1935 hours. 3/13/81 - 3/31/81 The turbine generator was synchronized on line_at 1350 hours. The reactor power was slowly increased and attained 90% full power on >brch 22, 1981 with the generator gross load at 845 i 10 IMe. The reactor power was maintained at 90% full power for the rest of the month. t y I l

v r March, 1981 I REFUELING INFORMATIO'{ DATE: 7

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 If answer is yes, what, specification chan2e or other license amendment?

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)? No Reload analysis is scheduled for completion as of December, 1981. technical specification changes or other license amendments identified to date. 2 S '. Scheduled date(s) for submitting proposed licensing action and supporting information. January, 1982

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.

None identified to date. f .

7. The number of fuel assemblics (a) in the core and (b) in the spent fuel storage pool.

44 - Spent Fuel Assemblics ! 177 (b)

  • 8 - Neu Fuel Assemblies (a)

! 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, i in number of fucl assemblics. 1 i Present 735 Increase size by 0 (zero)

9. The projected date of the last refueling that can be discharged to the spent fuci pool' assuming the present licensed capacity. '

\. Date 1988 (assuminF ability to unload the entire core into the spent fuel { pool.is maintained) l o 

                                    .        --,n.                        . .          ,   ,,,        -  - , - - - - , , . , ~

COMPLETED FACILITY CllANGE REOUESTS_ FCR No: 78-159 _SYSTi:M: Process and Area Radiation Monitoring COMPONENT: Various Radiation Monitors (see below) CllANGE, TEST OR EXPERIMENT: On October 18, 1979, the work as required by FCR 78-159 was completed. This FCR involved the replacement of the inboard and outboard bearings with Fafnir #203PP and #205PP sealed bearings in the pump notor of the following gaseous Radiation Monitors: RE 1003A RE 5052 RE 1003B RE 5327 RE 2024 RE 5328 RE 2025 RE 5403 RE 5029 RE 5405 . RE 5030 REASON FOR Tile CIIANGE: Due to the high ambient temperatures around the Radiation Monitors, motor bearing seizure has been recurrent. SAFETY EVALUATION: The replacement of the old bearings with scaled bearings wil1 improve the operation of the radiation monitors. In addition, the new bearings have been packed with a qualified high temp grease. The changes will enhance the operability of t.he monitors and will not adversely affect the safety of the plant. An unreviewed safety question does not exist. I 1

C0fiPLETED FACILITY CliANGE REQUEST FCR No: 79-031 SYSTEM: Portable Communications COMPONENT: N/A PROPOSED CHANCE, TEST, OR EXPERIMENT: On August 17, 1980 the installation of a portable cocmunications system uas completed. This system consists of, portable radios and a passive antenna system to enable emergency communications between the auxiliary building, containment, and the control room. REASON FOR THE CHANGE: This system was installed to comply with Toledo Edison's commitment to provide portable communication equipment in selected locations for emergency communications. This commitment was made in the Fire Hazard Analysis Report, Revision 2 (Table 4-1, Section DS, Sheet 26). SAFETY EVALUATION: This FCR is non-nuclear safety related except for the core drill / cutouts. Installation in accordance with the core drill report and PICA will pre-clude those portions from creating any new adverse environment. SAFETY EVALUATION _: An unreviewed safety question is not involved. 1 l 4 a

COMPLETED FACILITY CHANGE REQUESTS FCR No: 79-159 . SYSTEM: Containment Isolation - COMPONENT: NA CHANCE, TEST OR EXPERIMFNT_: FCR 79-159 has been implemented to provide admin-1strative controls and verification on the placement of caps on lines associated with containment isolation. The work portion of this FCR was completed on tby 21, 1979. . REASON FOR Tile CllANGE: This FCR was written to verify containment isolation on capped lines associated with the containment penetrations and to verify the integrity of systens that could be af fected by uncapped lines. SAFETY EVALUATION: FCR 79-159 provides verifiqation of capped pipes to ensure the establishment of containment isolation of capped lines associated with containment penetrations as well as verifies capped conditions on lines i. hat could affect the proper operation of certain safety related systems. The associated caps have been reviewed to ensure that they will not adversely affect system operation given precaution to preclude overpressurization of lines between closed valves and valve caps. Thic dacs not af fect any event described in the Safety Analysis Report nor does it affect the Station Technical Specifications. An unrevicued safety question is not involved, therefore, no license amendment. is required.

COMPLETED FACILITY CllANCE REQUESTS FCR N0_: 79-414 SYSTEM: Main Steam COMP 0 MENT _: Conduits 2-57112A and 2-57113A CllANGE, TEST OR EXPERIMENT: On September 3, 1980, the upgrading of four (4) supports to seismic class I was completed. These supports were installed Drawing for the essential cenduits 2-57112A and 2-57113A, located in Room 500. E-302A was revised to show the proper support details required f n the up-grading of the above listed conduits. REASON FOR Tile CIIANGE: It was determined from Nonconformance Report 14-79 that the supports for conduits 2-57112A and 2-57113A should be scismic class I. SAFETY EVALUATION: This FCR provided for the upgrading of four (4) supports to Seismic Class I. Installation in accordance with drawing E-302A will preclude creating any new adverse environments. An unreviewed safety question does not exist. o 4

COMPLETED FACILITY CHANGE REQUESTS FCR NO: 80-117 SYSTE'!: Ilydraulic Snubbers COMPONENT: Various _CIIANGE, TEST OR EXPERIMENT: On September 23, 1980, the work as required by FCR 80-117 was completed. This FCR involved changing the orientation of the reservoir tubing on various snubbers. The snubbers affected were: EBB-1-ll5 12482 EBB-1-ll5 12483 EBB-1-SR9 12474 . EBB-1-SR9 12480 EBB-1-SR8 12447 EBB-1-SR8 12449 REASON FOR TllE CHANCE: Previously, it was virtually impossible to remove or install the snubbers without getting air trapped in the lines. This modifica-tion will allow the proper hookup between the snubbers and reservoirs. SAFETY EVALUATION: The changes in the reservoir tubing orientation on the above listed snubbers will not change the function of the snubbers. The changes will provide added insurance that air will not be trapped in the tubing and adversely affect the snubber operation. On this basis, an unreviewed safety question does not exist. 1 l l s A

r o" COMPLETED FACILITY CHANGE RFQUESTS FCR NO: 80-273 SYSTEM: Process and Area Radiation Monitoring COMPONENT: RE 5029A and 5030A CHANGE, TEST OR E".PERIMENT: FCR 80-273 was written to decrease the sensitivity of Radiation Monitors RE5029A and RE5030A by at least a factor of 1000. This was accomplished by replacing the detectors in RE5029A and RE5030A with Victorcen detecto s, model number 843-20B, which have a 100:1 sensitivity reduction. The work, as required by this FCR, was completed January 10, 1981. REASON FOR THE CHANGE: In December of 1980, the containment particulate airborne monitors were found nearly off scale. Whan the readings are off scale, the instruments will be inoperable, and in accordance with T.S. 3.4.6.1.a, operation of the . cactor can only continue for 30 days. SAFETY EVALUATION: The subject radiation detectors RE5029A and RE5030A are utilized for monitoring of the containment during normal operation and for de-tection of containment radioactivity resulting from a reactor coolant pressure boundary (RCPB) leak. The Davis-Besse Unit 1 Technical Specifications address these monitors it Sections 3/4.3.3.1, 3/4.3.3.6, and 3/4.4.4.6.1. The requirements specify that the' measurement range be 10 to 106 cpm. There is no limit on sensitivity or response time. The effect of a reduction by a f actor of 100 in sensitivity has been evaluated. The sensitivity as noted below is below the maximum permissable conce'ntration (MPC) for a restricted area for activity in air, as specified in 10 CFR 20, Appendix B, Table 1, Column 1. This satisfies the statement in Section 11.4.2.2.5 of the FSAR that requires the ability to alarm at MPC. Monitor Sensitivity at Three (3) GPM Monitor Isotope of Interest Sensitivity (uc/cc) MPC (uc/cc) Particulate Ca l37 3 x 10-9 6 x 10-8 The NRC Safety Evaluation Report (SER) Supplement 1, Section 5.2.4, stated that the RCPB leakage detection systems "are generally in accordance with the re- , commendations of Regulatory Guide 1.45." This guide states that a one (1) gallon per minute (gp 2) leak rate should be detected within one hour. Calcula-tions indicate that monitors RE 5029A and RE 5030A will be capable of detecting a RCPB leak rate of one (1) gpm within one hour. Based on the above, it is concluded that the proposed reduction in sensitivity will not result in a change in the Technical Specifications incorporated in the license or an unreviewed safety question per the definition of 10 CFR 50.59.}}

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