ML19350A283

Monthly Operating Rept for Feb 1981

ML19350A283
Person / Time
Site:	Davis Besse
Issue date:	03/06/1981
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML19350A282	List: ML19350A280 ML19350A283
References
NUDOCS 8103130397
Download: ML19350A283 (12)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 W'IT Davis-Besse Unit 1 DATE March 6, 1981 COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, Ext.

251 MONTH Februarv 1981 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (31We-Net) (31We Net) 3 0 g7 885 2 0 18 885 3 37 39 876 4 ,

358 20 887 738 884 5 21 6 851 22 887 7 879 23 886 g 884 2.g 886

, 880 25 884 10 884 26 884 3g 867 27 884-875 884 12 28 13 885 29 14 882 30 15 886 33 16 885 INSTRUCTIONS On this format,hst the asera;e daily unit power leselin MWe-Net for each day in the reporting inonth. Compute to the nearest whole rnegawati.

(9/77i L

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE March 6 1981 CO3tPLETED BY Bilal Sarsour TELEPilONE 419-259-5000, Ext.

251 OPER ATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

2. Reporting Period: Februarv. 1981

3. Licensed Therm 21 Power (31Wt):

2772

4. Nameplate Rating (Gross SIWE): 925

5. Design Electrical Rating (Net 31We): 906

6. Staximum Dependable Capacity (Gross 31We): 934

7. Af aximum Dependable Capacity (Net 31We): 890

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

9. Power Lesel To Which Restricted. If Any (Net 31We):

10. Reasons For Restrictions,if Any:

This Sionth Yr to.Date Cumulatise 672 1,416 30,725

11. Ilours in Reporting Period

12. Number Of Hours Reactor Was Critical 619 753 15,137.2

13. Reactor Reserve Shutdown liours 0 0 2,882.1

14. Hours Generator On Line 608.1 736.0 13,783.8

15. Unit Resene Shutdown flours 0 0 1,731.4

16. Gross Thermal Energy Generated (StWil) 1,604,655 1,796,528 28,701,334 .

17. Gross Electrical Energy Generated IAlWill 540,284 ,,, 602.718 9.578.052

18. Net Electrical Energy Generated (31Wil) 512.173 563.902 8.828.403

19. Unit Senice Factor 90.5 52.0 45.5 i 20. Unit Availability Factor 90.5 52.0 51,5

21. Unit Capacity Factor (Using 11DC Net) - 85.6 44.7 34.2

22. Unit Capacity Factor (Using DER Net) 84.1 __

44 33.6

23. Unit Forced Outage Rate 9.5 48.0 26.4

24. Shutdowns Scheduled Over Nest 6 Stonths (Ty pe. Date.and Duration of Each t:

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

l 23. Units In Test Status IP.ior to Commercial Operationi: Force::st Achiesed

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l INITIA L CRITICALITY ISITIAL ELECTRICITY CONINIERCIAL OPERATION (4/77)

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DOCKET NO. 50-346 UNIT SHUTDOWNS AND POW .R REDUC 110NS UNIT NAME Davis-Besse Unit 1 ,

DATE March 6 1981 i COMPLETED BY Bilal Sarsour February. 1981 REPORT MOlTHI TELEPIIONE 419-?59-5000- Ext. 251  ;

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,{g 3 yYh Licensee ,Ev, gn, Cause & Corrective ,

No. Date

g. 5g 4 .3 s s Event 3,1 91 Action to H y: $ jgy Report a mu yV Prevent Recurrence 5

63.9 NP-33-81-01 CB PUMPXX The unit was shutdown to repair the 1 81 01 06 F A 1 seals in the Reactor Coolant Pumps Continued -

1-2 and 2-1.

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3 4 I 2 Exhibit G. Instructions F: Forced Reason: Method:

1-Manual for Prep.tration of Data S: Schedu!cd A Equipment Failure (Explain) 2 Manual Scram. Entry Sheets for Licensee

n. Maintenance or Test Event Reporn (LER) File (NUREG-C. Refueling 3 Automatic Scram.

D-Regulatory Restriction 4@heHf W b 0161)

I'.-Operator Training & License Examination 4-Continuation S

F Administrative S-Reduction "# " " ' ' '

G Operational Eisor Ilixplain) 6-Other (9/77) Il-Ot her ( E xplain)

OPERATIONAL SU10!ARY February, 1981 The unit shutdown which was initiated on January 6, 1981 because of problems with the seals in the Reactor Coolant Pumps 1-2 and 2-1 continued until 1557 hours0.018 days <br />0.433 hours <br />0.00257 weeks <br />5.924385e-4 months <br /> on February 3. 1981 when the turbine generator was synchronized.

Reactor power was increased and 100% full power was achieved on February 7, 1981.

Reactor power was held at between 99% and 100% of full power for the rest of the month with the turbine generator gross load at approximately 925 + 10 IMe.

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DATE:

February, 19S1 REFUELING INFORMATION Na=e of f acility: Davis-Besse Nuclear Power Station Un'it 1 1.

February. 1982

2. Scheduled date for next refueling shutdown:

May, 1982

3. Scheduled date for restart following refueling:

4. Will refueling or resu=ption of operation thereaf ter requireisa yes, If answer technical what, specificarica change or other license a=end=ent?If answer is no, has the reload fuel de in general, will these be? Safety Review Cc=mittee and core configuration been reviewed by your, Plant to determine whether any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?

Reload analysis is scheduled for co=pletion as of December. 1981. No technical specification changes or other license atend=ents identified to date.

5. Scheduled date(s) for sub=itting proposed licensing actien and supporting inforcation. Januarv. 1992 Important licensing considerations associated with refueling, 6.

different e.g., new orfuel design or supplier, unreviewed design or pe methods, significant changes in fuel design, new operating procedures.

None identified te date.

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7. The nu=ber of fuel asse=blies (a) in the core and (b) in the spent fuel storage pool. ' 44 - Spent Fuel Assemblies 177 (b) 8 - New 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, in number of fuel assemblies.

Increase size by 0 (:ero)

Present 735

9. The projected date of the last refueling that can be discharged to the spent .

fuel pool assuming the present licensed capacity.

Date 1988 (assuning abilitv to unicad the entire core into the A-en* c mt pool is maintained) ,

COMPLETED FACILITY CHA*!GE REQUESI FCR NO: 77-412 SYSTEM: Hydrogen Dilution System COMPONENT: Blower Separator PI 5147 CHANGE, TEST, OR EXPERIMENT: FCR 77-413 was written to revise the Johnson Service Company drawing number I/F PI-5147 to reflect the relocation of stainless steel tub-ing from PI-5147 to its shutoff valve. Tubing supports for PI-5147 were installed September 24, 1980 per Supplement 1.

REASON FOR CHA' IGE: The relocation of this tubing was necessary to permit the instal-lation of a ladder in the #4 penetration room. Drawing I/F PI-5147 was then revised to reflect these changes.

SAFETY EVALUATION: The rerouting of stainless steel tubing from PI-5147 to the shut-off valve for PI-5147 did not affect the safety function of the hydrogen dilution system.

This change does not constitute an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUES_T FCR NO: 90-078 .

SYSTEM: Makeup and Letdown System COMPONENT: Letdown Line Stop Valve, HV-MU3 CHANQE, TEST, OR EXPERIMENr: On >by 28, 1980, the work portion of FCR 80-078 was completed. This FCR proposed eliminating the high letdoun temperature interlock (TS-MU8) which closes valve HV-MU3, and changing the setpoint on TSH-3745B from 160 F to 135 F.

REASON FOR CHANGE: Previously an instrument power supply failure on the YBU bus or in the NNI-X 9 (+ or - 24 VDC) supply would cause TS-MU8 to fail in the high direc-tion, and consequently close valve MU-3. The above change eliminates this inadver-tent valve closure on power supply failures.

SAFETY EVALUATION: The high letdown temperature interlock (TS-MU8) was originally designed to prevent damage to the purification demineralizers from high temperature letdown water. The valve closure was designed to occur at 135 F letdown water temperature downstream of flow element FE-MU7. This interlock is non-safety grade.

At present, there are two safety grade interlocks which also cause letdown isolation on high temperature. TSH 3745B closes valve MU2B and TSH 3745A closes valve MU1A and MUlB. These interlocks provide backup to the non-safety grade interlock.

With the change of setpoint on TSH-3745B from 160o to 135 F, letdown isolation will still be achieved at 135 F by closing valve MU33. Moreover, this temperature is measured at the inlet of the delay coil instead of downstream of flow element FE-MU7.

The setpoint on TSH-3745A is to remain at 160 F which will provide a backup to the safety grade 1350F interloch. The high temperature alarm T715 will still be actuated at 130 F. Also, the SFAS incident level 2 closure of valve MU3 to achieve contain-ment isolation will be unaffected by the above change.

Based on the above, it is concluded that the changes proposed by this FCR do not involve an unreviewed safety question.

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COMPLETED FACILITY CIIANGE REQUEST FCR NO: 80-126 .

SYSTDi: Safety Features Actuation System COMPONENT: V-5009 CHANGE, TEST, OR EXPERDIDiT: FCR 80-126 has been implemented to revise drawing E 58 13 Sheet 7A, Revision 3. This drawing was revised to show F-133 feeding V-5009 and not F114 as shoe in the previous revisions.

REASON FOR CHANGE: This drawing was revised to reflect the "as built" conditions for fusing V-5009.

SAFETY EVALUATION: This FCR involves only a drawing revision to reflect "as built" conditions and therefore an unreviewed safety question docs not exist, i

COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-161 SYSTEM: Steam Generator System COMPONENT: LT SP9Al CHANCE, TEST, OR EXPERIMENT: FCR 80-161 has been written to revise Bechtel's drawing Vendor Print No. 7749-M-329-37-3 (Johnson Service Conpany Drawing I/F LT SP9Al-A5, Sheet 4 of 18) to show a 6000# SA-105N - 1" x 1/2" Reducer which was installed on the icw pressure connector for LT SP9Al at the isolation root valve found on the 585' 9/16" elevation level.

REASON FOR CHANGE: When valve SP9Al was replaced per Maintenance Work Order 79-1591, a 6000#, SA105N - 1" x 1/2" reducer was installed, the Johnson Service Conpan/ drawing showed a 3000# SA105 PSN - 1" x 1/2" reducer uns to be installed.

Nonconformance Report 149-80 was issued to approve of the use of the 6000# SA105N 1" x 1/2" reducer and to revise the appropriate drawings to show the as-built condi-tions. Therefore, this drawing update is required for final resolution of NCR 149-80.

SAFETY EVALUATION: The installed change from a 3000# fitting to a 6000# fitting did not degrade the original installation, nor did it involve an unreviewed safety ques-tion or technical specification change.

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COMPLETED FACILITY CHANGE REQUEST _

FCR NO: 80-190 .

SYSTEM: Reactor Coolant Systen COMPONENT: Hangers PS-H7C, and 30-GCC-8-H5 CHANCE,~ TEST, OR EXPERIMENT: On August 29, 1980, the work package for FCR 80-190 was completed. This FCR implemented the modification of hangers PS-H7 (pressuri-zer spray).and 30-GCC-8-H5 (pressurizer relief) to' lengthen the piston settings of these-snubbers.

REASON FOR CHA' GE: These snubbers were determined to have piston settings outside

- the acceptable range as set forth in the Bechtel drawing number 12501-M-618.

SAFETY EVALUATION: This FCR involved the modification of two hangers to lengthen the piston settings of two suchbers such that their cold piston settings are within the acceptable range. This modification will not create any adverse environment

- conditions. Furthermore, the operability of these snubbers will be enhanced by thic change. An unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-200 SYSTEM: Main Steam COMPONENT: Hanger 3A-EBD-19-H10 CHANGE, TEST, OR EXPERIME'iT: On September 3, 1980, the modifications to hanger 3A-EBD-19-H10 (main steam) were completed. This modification involved lengthening the piston setting of the snubber.

REASON FOR CHANGE: This snubber was determined to have its piston setting cacside of the acceptable range as set forth in the Bechtel drawing number 12501-3-618.

SAFETY EVALUATION: 'This FCR modified hanger 3A-EBD-19-H10 to increase the snubber piston setting to a value within the acceptable range. Increasing the piston setting in this manner will enhance this hanger's ability to perform its intended safety function. No new adverse environment will be created by this modification.

An unreviewed safety questien is not involved.

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