ML19210C486

Monthly Operating Rept for Oct 1979

ML19210C486
Person / Time
Site:	Davis Besse
Issue date:	11/09/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19210C484	List: ML19210C483 ML19210C485 ML19210C486
References
NUDOCS 7911140332
Download: ML19210C486 (10)

Text

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AVERAt;E DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

Davis-Besse Unit 1 3.lT November 9, 1979 DATE Erdal Caba C051PLETED BY (419) 259-5000, TELEPHONE Ext. 236 5!ONTil October, 1979 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE D AILY POWER LEVEL (51We Net)

(51We Net) 879 0

37 0

882 18 2

0 3

883 39 4

884

o,

0 558 80 5

yg 6

0 22 7'6 653 0

23 7

641 647-24 8

9 877 25 316 0

10 883 26 0

887 27 II 0

885 23 12 0

13 886 29 0

785 30 14 0

15 31 0

16 INSTRUCTIONS On this format. list the average daily unit power levelin S!We Net for each day in the reporting inonth. Compute to the nearest whole megawatt.

(9/7.7 i 7911140 3 3 d 1321 341

OPERATING DATA REPORT DOCKET NO.

50-346 NV Der 9' 1979 DATE COMPLETED BY Erdal Caba TELEP110NE 419-25 9-5000, Ext.

236 OPERATING STATUS Notes Davis-Besse Unit 1

1. Unit Name:

October, 1979

2. Reportin; Period:

2772

3. Licensed Thermal Power i.\\lWt1:

925

4. Nameplate Rating tGross 31Weg.

906

5. Design Electrical Rating tNet MWe):

6. 51aximum Dependable Capacity (Gross 31We): to be determined to be deter'ined

7. 5!aximum Dependable Capacity (Net 3the):

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

N ne

9. Power Level To %hich Restricted,if Any (Net.\\lWe):

10. Reasons For Restrictions. If Any:

This Slonth Yr. to-Date Cumulatise 745 7.294 19,061 II. Ifours In Reporting Period 419.1 4.043.5 10.675.3

12. Number Of flours Reactor Was Critical 142.2 2,085.5 2,875.8

13. Reactor Reserve Shutdown Hours 397.6 3,900.6 9,633.8

14. Hours Generator On-Line 0

1,728.2 1,728.2

15. Unit Resene Shutdown Hours 950,616 9,601,764 19,789,334

16. Gross Thermal Energy Generated (MWH) 320,890 3,202.078 6,385,833

17. Gross Electrical Energy Generated (51% H) 294,028 3,008,035 6,049,495

18. Net Electrical Energy Generated (MhH) 53.4 53.5 52.3

19. Unit Senice Factor 53.4 77.1 62.4

20. Unit Asailability Factor to be determined

21. Unit Capacity Factor iUsin; 31DC Nett 43.6 45.5 38.5

22. Unit Capacity Factor tUsing DER Net) 42.8 10.1 20.4

23. Unit Forced Outage Rate

24. Shutdowns Se eduled Oser Nest 6 Months (Type. Date.and Duration of Eacht:

b Refueling outage to start March 15, 1980.

November 12, 1979

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

26. Units In Test Status (Prior to Commercial Operationi:

Forecast Achiesed INITIAL CRITICALITY INITIA L ELECTRICITY COMMERCIAL OPERATION 1321 342 (4/77)

D UNIT S!!UIDOWNS AND f OWE!t REDUCI3ONS d

v

--Besse Unit 1 Ui DATE November 9, 1979 COStP!. Ell:D BY l'rd a I Caba REPORT MON MI October. 1979 IJE IONE 419-259-5000. Ext. 236 w.

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16 79 10 5 S

48.8 A

1 NA CJ VAINEX Shutdown to repair pressurizer spray valve RC 2.

17 79 10 15 F

142.2 A

3 NP-32-79-11 IIA INSTRU Capacitor failure in Integrated Con-trol System (ICS) pulser circuit to the turbine electro-hydraulic control system.

Refer to attached summary i

for further details.

18 79 10 25 F

156.4 A

3 NP-33-79-121 CB UKTBRK Loss of Reactor Coolant Pump 2-2 from blown fuse in the DC power supply starting a pump two minute *;1me delay trip relay with Reactor Coolant Pump 1-1 already shutdown.

I 2

3 4

F: Forced Reason:

Method:

Exhibit G -Instructions S: Schedu!ed A Equipment Failure (Explain) 141anual for Preparation of Data B-Maintenance oilest 241an ual Scram.

I'ntry Sheets for Licensee 3 Antomatic Sciam.

Event 1(eport (LI:lt)i ile (NUREG-C Refnelmg D Regulatory Restriction 4-Ot he: (lisplam)

Ol td )

1: Operaio Training & License Examinatio:.

5 F Adnunntiiine L shibit !- Same Sonrec G Opesanonal I nor (ihplain)

U (9/77) 11 Other (I splam) c::=

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OPERATIONAL

SUMMARY

FOR OCTOBER, 1979 10/1/79 - 10/5/79 Reactor power was maintained at 100 percent until 2025 hours0.0234 days <br />0.563 hours <br />0.00335 weeks <br />7.705125e-4 months <br /> on October 5, 1979 when the reactor was manually shutdown to repair the pressurizer spray valve RC2.

10/6/79 The unit remained shutdown to repair the pressurizer spray valve.

10/7/79 The turbine generator was synchronized at 2115 hours0.0245 days <br />0.588 hours <br />0.0035 weeks <br />8.047575e-4 months <br /> and reactor of full power was increased and maintained between 98-99 percent power.

10/11/79 - 10/14/79 The unit operated at approximately 100 percent full power uith the turbine-generator gross load at 920 + 5 XJe.

At 1678 hours0.0194 days <br />0.466 hours <br />0.00277 weeks <br />6.38479e-4 months <br /> on October 14, 1979, group 7 rod 7 API was declared inoperable and reactor power was reduced to approximately 59 percent.

10/15/79 At approximately 1227 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.668735e-4 months <br /> on October 1:,1979, a capacitor f ailed in the Integrated Contro' System (I~S) pulser circuit o

the turbine electro-hydraulic

.atrol system. This capacitor failure caused che turbine control valves to open which 1cwered the main stean line header pressure. The ICS responded to the low header pressure by increasing both reactor power and feed-water which resulted in a reactor protection system reactor trip at the reduced high flux setpoint of approximately 68.8 percent.

of full power.

While reclosing the generator output breaker 34560 at approxi-mately 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br />, "J" bus tripped which resulted in a de-energization of the startup transformer 01 and a station loss of off-site power. Both emergency diesel generators automati-cally started.

The Steam and Feedwater Rupture Control System (SFRCS) actuated from the loss of all four reactor coolant pumps, both auxiliary feed pumps started and natural circulation was established in the Reactor Coolant System (RCS). The cause of the station loss of off-site power was a blowing out of the internals of the nuf fler on generator output breaker 34560 when it opened for the trip which caused a fault to ground when the breaker was reclosed.

10/16/79 - 10/22/79 The unit renained shutdown until October 31, 1979 when the tur-bine-generator was synchronized.

Reactor power was increased to 92 percent of full power at approximately 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on October 22, 1979 for the xenon equilibrium hold.

10/23/79 Reactor power was decreased to 70 percent of full power when Reactor Coolant Pump 1-1 seal destaged due to first and third stage seal failure. The reactor coolant pump was tripped at 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />.

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~

OPERATIO:IAL SEKMARY FOR OCTOBER,1979 PAGE 2 10/24/79 - 10/25/79 Reactor power was maintained at approximately 70 percent with three reactor coolant pumps in operation on October 25, 1979, when a reactor trip occurred.

At approximately 1226 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.66493e-4 months <br /> on October 25, 1979, station per-sonnel de-energized E5 bus to remove station transformer ST 1 f rom service. At approximately 1238 hours0.0143 days <br />0.344 hours <br />0.00205 weeks <br />4.71059e-4 months <br />, Reactor Coolant kamp 2-2 tripped from a blown fuse in the DC power supply. The tripping of this pump started a two minute time delay trip relay with Reactor Coolant Pump 1-1 already shutdown because of seal staging difficulties.

The Reactor Protection System tripped the reactor on a " flux to nu=ber of reactor coolant pumps" actuation.

A design deficiency was discovered in the reactor coolant pump interlock circuit which has been corrected with the addition of surge suporessors.

10/26/79 - 10/31/79 The unit remained shutdown to replace the seals of Reactor Coolant Pump 1-1.

It was later decided to replace the seals of all four reactor coolant pumps when it was determined that the station loss of of f-site power incident on October 15, 1979 had an adverse effect on the reactor coolant pump seal perfor-mance.

G 1321 345

REFUELI1?C INFORMATI0'T DATE:

October, 1979

[

Davis-Besse Nuclear Power Station Unit 1 1.

Name of facility:

March, 1980 2.

Scheduled date for next refueling shutdown:

May, 1980 3.

Scheduled date for restart following refueling:

Will refueling or rese=, tion of operation thereafter require a technical 4.

If answer is yes, what, specificatien change or other license amend ent?

in general, will these be? If answer is no, has the reload fuel design and core configuratica been reviewed by your plant Safety Review Cctaittee to determine whether any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached Scheduled date(s) for sub=itting proposed licensing action and supporting 5.

inf o r=a tion. -

December. 1979 licensing considerations associated with refueling, e.g., new or 6.

Important fuel design or supplier, unreviewed design or performance analysis different methods, significant changes in fuel design,.new operating procedures.

fuel pool capacity expansion program was approved by the NRC in The spent Amendment 19 to the operating license received August 1, 1979.

~

7.

The number of fuel asse=blies (a) in the core and (b) in the spent fuci storage pool.

0 (zero)

(a) 177 (b)

The present licensed spent fuel pool storage. capacity and the size of any 8.

increase in licensed storage capacity that has been requested or is planned, in number of fuel assemblies.

475 (735 total) 260 Increase size by Present The projected date of the last refueling that can be discharged to the spent 9.

(

fuel pool assuming the present licensed capacity.

1989 (assu=ing ability to unload the entire core into the spent Date fuel pool is maintained and the unit goes to an 18 month refueling cycle)

i REFUELEIG DEOP3.ATION (Continued)

OCTOBER. 1979 i+

Pagr 2 of 2 4.

The following Technical Specifications (Part A) vill require revision:

2.1.1 & 2.1.2 - Reactor Core Saf ety Limits (and Bases) 2.2.1 - Reactor Protection System Instrunentation Setpoints a

(and Bases) 3.1.3.6 - Regulating Rod Insertion Limits 3.1.3.7 - Rod Program 3.2.1 - Axial Power Imbalance (and Bases)

The following Technical Specifications (Part A) =ay also require revision:

3.1.2.8 & 3.1.2.9 - Borsted k'ater Sources (and Esses) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - DNB Par =e.ters (and Bases)

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e 1321 347 e

e 8

FACILITY CHANGE REOUEST COMPLETSD DURING OCTOBER, 1979 FCR NO:

77-478 SYSTEM: Auxiliary Feedwater (AFW)

COMPONENT: AFW Pump speed control transfer switch position indication CHANCE, TEST. OR EXPERI:!ENT: On July 11, 1977 station instrument and control tech-nicians found that the wiring of the position indicating lights for the AFW pump speed control local / remote transfer switch was incorrect as was the applicable "as-built" drawings.

The circuitry was rewired to make the indicatirg lights operate properly. Revisions to Bechtel drawings E-453 Sheet 1, t-289 Sheet 1, and E-298 Sheet 1 were completed by the unit architect / engineer, Bechtel Company, in order to document the corrected "as-built" wiring configuration.

REASON FOR THE FCR:

FCR 77-478 was written to docu=ent the corrected wiring config-uration of this indicating lamp circuit.

SAFETY EVALL'ATION: This FCR involves the rewiring of contacts in the auxiliary feed pump turbine speed control systen in order to have the indicating lights associated with the local / remote transfer switch properly reflect the position of the switches.

This change enhances operation of the systen and does not result in an unreviewed safety question.

1 321 348

E FACILITY CHANGE REQUEST COMPLETED DURING OCTOBER, 1979 FCR NO: 78-521 SYSTEM: Post Accident Radiation Monitors COMPONENT: Sampling Pumps CHANCE, TEST, OR EXPERIMENT: On September 5,1979 work was completed for FCR 78-521.

This FCR was written to investigate the possibility of reducing

• he speed of the a.r samplir.g pumps of RE3029 and RE5030, the Containment Post Accident Radiation Monitors.

Af ter analysis, the unit a chitect/ engineer, Bechtel Corporation determined that the function of these monitors would not be affected by this change.

The pump pulley size and the size of the driving belts were changed to reduce the samp'ing rate from approxi-mately 8.5 CFM to 4 CFM.

Also new flow meters were installed in order to properly monitor the lower flow rate. These changes were made with the concurrence of both Bechtel Corporation and the monitor vendor, Victoreen Incorporated.

REASON FOR THE FCR: This reduction in pump speed was undertaken to decrease pump wear and internal heating as well as to decrease the load on the pump drive motors in an attempt to reduce the frequency of motor and p mp bearing failures (see Licensee Event Reports NP-33-79-93, NP-33-79-95, NP-33-79-42, NP-33-79-37, NP-33-78-143, NP-33-78-127, NP-33-78-111, SP-33-78-105, NP-33-73-101, NP-33-78-77, NP-33-78-54, NP-33-78-45, anc' 'IP-33-78-30). These two radiation monitors are the only ones which have experienced repetitive tearing failures.

This is attributed to the fact that these particular monitors are located in areas where the ambicat temperature during plant operation is high (mechanical penetration rooms).

SAFETY EVALUATION: The subject radiation detectors RE5029A, B, C and RE 5030A, B, and C are utilized for post-accident r.onitoring and monitoring of the containment during normal operation.

For the post-accident function, the gaseous monitors are needed. For detection of containment radioactivity resulting from a reactor coolant pressure boundary leak, the monitors of interest are the particulate and gaseous monitors.

The Davis-Besse Unit 1 Technical Specifications address these monitors in Sections 3/4.3.3.1, 3/4.3.3.6, 3/4.4.6.1 and 3/4.4.6.2.

The only requirements imposed deal with the particulate and gaseous radioactivity monitors and specifv that the measure-ment range be 10 to 106 cym.

There is no 1Lnit on sensitivity or t sponse time. The iodine radioactivity monitors are not required.

The effect of the proposed reduction in blower flow rate to approximately four (4)

CFM has been evaluated. The expected sensitivicies for approximately three (3) CFM as noted below are in the rar,e of the values specified in the FSAR which were based in 8.5 CFM and are well b;10w the maximum permiss.'.ble concentration (MPC) for a res-tricted 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 re-quires the ability to measure MPC.

Cs

FACILITY CHANGE REQUEST COMPLETED DURING OCTOBER, 1979 FCR 78-521 PACE 2 Monitor Sensitivity at Three (3) CFM Monitor Isotope of Interest Sensitivity (pc/cc)

MPC (pc/cc) 137 3 x 10-11 6 x 10-8 Particulate Cs Gaseous Xe 3 x 10-7 1 x 10-5 133 Iodine 1131 2 x 10-12 9 x 10-9 The NRC Safety Evaluation Report (SER) Supplement 1, Section 5.2.4 stated that the leakage detection systems "are generally in accordance with the recocmencations of (1) gallon per minute (gpm)

Regulatory Guide 1.45."

This guide states that a one leak rate should be cet;cted within one hour. The options for detection include the containment particciate and gaseous radioactivity monitors. Calculations indic. ate that these monitors will be capable of detecting a leak rate of one (1) gpm within one hour, utilizing a blower flow rate of three (3) cfm.

This is true whether the containment is being purged or not.

Based on the above, it is concluded that the proposed reduction in blower flow rate to approximately three (3) to four (4) cfm will not result in a change in the Tech-nical Specifications incorporated in the license or an unreviewed safety question per the definition of 10 CFR 50.59.

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