ML19270G651

Monthly Operating Rept for May 1979

ML19270G651
Person / Time
Site:	Davis Besse
Issue date:	06/08/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19270G650	List: ML19270G649 ML19270G651
References
NUDOCS 7906140408
Download: ML19270G651 (11)

Text

AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

Davis-Besse Unit 1 June 8, 1979 DATE Erdal C. Caba COMPLETED BY 419-259-5000 Ext.

TELEPilONE 236 MON I May, 1979 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAll 3WER LEVEL (MWe-Net)

(MWs u) 0 1

0 g7 0

0 2

gg 3

0 19 0

O 4

0 20 0

0 5

y, 0

0 6

22 0

0 73 7

0 0

8

.24 0

0 9

25 0

10 0

26 0

1I O

27 0

0 12 28 0

13 29 0

14 30 0

IS 0

3g 0

16 INSTRUCTIONS On this format, list the average daily unit power levelin MWe Net for each day in the reporting month. Compute to the nearest whole megawatt.

(0177) 7 006140 L/d g

OPERATING DATA REPORT

~

50-346 DOCKET NO.

DATE June 8, 1979 Erdal G. Caba COh!PLETED BY TELEP 'ONE 419-259-5000, Ext.

~

236 OPERATING STATt!S Notes Davis-Besse Unit 1

1. Unit Ngme:

May, 1979

2. Reporting Period:

2772

3. Licensed Thermal Power (31Wt):

925

4. Nameplate Rating (Gross Slwel:

906

5. Design Electrical Ratin;INet SlWe):

to be detemined

6. 51aximum Dependable Capacity (Gross alwe): to be detemined

7. Staximum Dependable Capacity (Net 31We):

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

Zero

9. Power Level To which Restricted.If Any (Net 31We):

NRC OIE Bulletins and Shutdown Orders

10. Reasons For Restrictions.lf Any:

This hionth Yr.-to Date Cumulative 744 3,623 15,388 II. Ilours in Reporting Period O

1,747.4 8,37.9.2

12. Number Of flours Reactor Was Critica!

744 892.2 1,682.5

13. Reactor Resene Shutdown liours 0

1,675.1 7,408.3

14. Ilours Generator On-Line 744 744 744

15. Unit Rewne Shutdown flours

16. Gross Thermal Energy Generated (51Wil) 0 3.879.097 14.066.667

17. Gross Electrical Energy Generated 13th H) 0 1.293.268 4.677.023 0

1.212.558 4.254.018

18. Net Electrical Energy Generated (51WH) 0 46.2 50.0

19. Unit Senice Factor 100 66.8 55.6

20. Unit Asailability Factor

21. Unit Capacity Factor (Using S!DC Net)

_ to be detemined 0

36.9 34.3

22. Unit Capacity Factor (Using DER Net) 0 4.8 22.4

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Next 6 alonths (Type.Date.and Duration of Eacht:

June 17, 1979

25. If Shut Down At End Of Report Period. Estimated Date et startup:

26. Units in Test Status (Prior to Commercial Operation):

Forecast Achiesed INITIAL CRITICALITY INITIA L ELECTRICITY CONIN!ERCIAL OPERATION (9/77)

?

1 DOCKETNO. 50-346 Davis-Besse Unit 1 UNIT S!!UIDOWNS AND POWLit REDUCTIONS UNIT NAME DATE COMPLETED BY ch,r1 ne n A1, REPORT MONTil TELEPl!ONE 419-259-5000. Ext. 251 e

.$?

3

.Y Licensee Eg Cause & Corrective No.

Date g

gg g

,g g 5 Event s?

93 Action to Report

• N0 8'

Prevent Recurrence H

Ji 5 =g u

e C

9 79 03 30 S

744 D

1 N/A N/A N/A The unit remaided in an outage thh T

entire month. Refer to the attached Q

Outage Summary of May 1979 for outage g activities this mon'th.

g C.

N W"

l*"

3 4

I 2

Method:

Exhibit G. Instructions F: Forced Reason:

l. Manual for Picparation of Data S: Schedu!cd A igde.nent Failure (Explain) 2-Manual Scram.

Entry Sheets for Licensee B-Maintenance ci Test 3 Automatic Scram.

Event Report (LER) File (NUREG.

qd C-Refuelmg I

D Itegulatory itestriction 4-Other (Explain) 0161)

Q l'. Operatur Training & License Examination 5

F Adminntrative Exhibit I. Same Source G-Operational E ror (lixplain) t (9/77)

Il-Other (Explain)

OUTACE SU> DIARY May, 1979 The unit outage which began at 2142 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.15031e-4 months <br /> on March 30, 1979 was still in progress through the end of thy,1979.

The outage was extended longer than anticipated to respond to NRC IE Bulletins 79-05,79-05A, and 79-05B. There have also beec. addi-tional NRC startup restraints which were imposed as a result of an ongoing analysis of the Three Mile Island incident.

The following are the more significant outage activities performed during the month of thy:

Procedure modificatkon to comply with NRC requests.

In addition, modifi-1.

cations of procedures were made due to the re-evaluation of the Babcock and Wilcox small break analysis.

2.

Personnel instruction on what the procedure major modifications involved.

3.

The installation of an additional anticipatory reactor trip system (Faci-lity Change Request 79-176).

4.

The installation of auxiliary feedwater flow indication on the Auxiliary Feedwater System. Also testing was performed on this system to resolve NRC commitments.

5.

The performance of eighteen month surveillance tests which were required to be completed prior to the reactor internal vent valve test. This was done to lengthen the available operation time af ter this outage.

6.

The replacement of the Reactor Coolant Pump l-1-1 seals.

7.

The performance of hydraulic snubber surveillance testing which.is still in progress.

8.

Testing was performed on both Moisture Separator Reheaters.

The Moisture Separator Reheater 2 was inspected and several instrument tubes were plugged and one broken instrument tube was repaired.

324

.EFUELING INFORMATION DATE:

May, 1979 Davis-Besse Nuclear Power Station Unit 1 1.

Name of facility:

2.

Scheduled date for next refueling shutdown:

March, 1980 3.

Scheduled date for restart following refueling:

May, 1980 4.

Will refueling or resumptien 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 re.oad 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)?

Yes, see attached 5.

Scheduled date(s) for submitting proposed licensing action and supporting information.

December, 1979 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 fuel design, new operating procedures.

The spent fuel pool capacity expansion program is awaiting a final NRC Safety Evaluation Report to proceed in time for completion prior to refueling. All licensee submittals are complete including environmental assessment questions.

7.

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

(a) 177 (b) 0 (zero) 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.

475 (735 total)

Present 260 Increase size by 9.

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

March, 1980 - May, 1980 (assuming ability to unload the entire core into Date the spent fuel pool is maintained. )

3 I

e REFUELING INFORMATION Cont'd May, 1979 Page 2 of 2 4.

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

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System Instrunentation Setpoints (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) may also require revision:

3.1.2.8 6 3.1.2.9 - Borated Water Sources (and Bases) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - DNB Parameters (and Bases) 9

)

FACILITY CHANGE REQUESTS COMPLETED DURIt'G MAY, 1979

.FCR No.77-475 SYSTEM: Emergency Diesel Generator 1-2 COMPONENT:

Control and Relay Panel C3616 CHANGE, TEST, OR EXPERDIENT: Facility Change Request (FCR)77-475 was written to revise General Electric Drawing 0132D4596, Sheet 2, to correct an error:in the draw-ing. Thi's drawing is the connection diagram for the Emergency Diesel Cencrator 1-2 local control panel. The specific change was to revise the external annunciator connections from terminals NN1 and NN2 to NN2 and NN3.

REASON FOR THE FCR:

It was found that incorrect wiring terminations had been made on terminal block NN in the diesel engine alarm cabinet because of the error on the General Electric drawing.

The -terminations were corrected and the local annunciator tested under Maintenance Work Order (MWO) 77-1037 on June 21, 1977. This correction is documented in General Electric Drawing 0132D4596, Sheet 2, Revision B1, dated February 8, 1978.

SAFETY EVALUATION:

This work, which corrects the vendor drawing for the external annunciator connections, completes the proper annunciator terminations to terminal block NN.

The change will allow the annunciator to function properly. This is not an unreviewed saf ety question.

FACILITY CHANGE REQUESTS COMPLETED DURING MAY,1979 FCR NO:

78-001 SYSTEM:

Emergency Ventilation System (EVS)

COMPONENT: Valves CV5024 and CV5025 (isolation valves between auxiliary building ventilation systems and EVS)

CHANCE, TEST, OR EXPERIMENT! FCR 78-001 was written to request the revision of Bechtel Piping and Instrument Diagram (P&lD) M-029A, "HV and AC Air Flow Diagram for Containment and Penetration Rooms", to reficct the as built conditions. The change made involved interchanging labeling on the controls and interlocks for valves HV2024 and HV2025 on the P&ID. This change was incorporated in Revision 22 of P&ID M-029A, dated February 16, 1978, by Bechtel, the unit architect / engineer.

REASON FOR THE FCR:

It was found that the labeling of the valves was interchanged, and the wiring of the controls and interlocks were interchanged between the valves.

The valves are connected in series in the flowpath.

Rather than actually rewiring the valves and associated controls. to correct this construction error, which would have required extensive plant modifications, a more acceptabic solution was to re-label the valves and associated controls on the P&ID.

SAFETY EVALUATION: The only change required on this FCR is to correct P&ID M-029A to reficct as-built locations of valves CV5024 and CV5025 and associated controls and interlocks. Toledo Edison Power Engineering Department has evaluated and con-firmed that the function of the EVS will not be affected by this change s v'2 s

Rev. 1 FACILITY CHANGE REQUESTS COMPLETED DURING DECE'IBER, 1978 FCR NO: 78-013 SYSTEM:

Final Safety Analysis Report (FSAR)

COMPONENT:

Chapter 14 CHANGE, TEST, OR EXPERIMENT:

FCR 78-013 was written on January 6, 1978, to nullify the erroneous requirement in the FSAR abstract of TP 800.25, " Shutdown From Outside the Control Room", that the' reactor be tripped from outside the Control Room as a startup test. However, the NRC requested that tha reactor be tripped from outside 1

of the Cont rol room as the FSAR abstract had originally required. The NRC request was in spite of the fact that tripping the reactor from outside the Control Room was inconsistent with the remainder of the FSAR. To comply with the NRC request, Test Procedure TP 800.25 was conducted on January 14, 1979 by tripping the reactor from outside the Control Room using the control rod drive breakers.

REASON FOR THE FCR: All of the accident analyses in the FSAR and the Davis-Besse Unit 1 Fire Hazards Analysis Report assume the reactor to be tripped prior to evacuation of the Control Room.

SAFETY EVALUATION: Administrative Procedures require that the reactor be tripped prior to evacuation of the Control Room.

The FSAR abstract of TP 800.25 requires that the reactor be tripped from outside the Control Room as a startup test.

This commitment was an error and is not consistent with the remainder of the FSAR; reference the Chapter 7 discussion of the Auxiliary Shutdown Panel.

This change in the Startup Test Program from that described in Chapter 12 of the FSAR does not therefore represent an unresolved or unreviewed safety question.

Conducting the test by tripping the reactor from outside the Control Room was done 1

to satisfy NRC desires and was done with prior NRC knowledge and approval.

3

FACILITY C'ANCE REQUESTS COMPLETED DURING MAY, 1979 FCR NO:

78-317 SYSTEM: Decay Heat Removal (DHR)

COMPONENT: Motor Operated Valves (MOVs) DH 2733 and DH 2734 CHANGE, TEST, OR EXPERIMENT:

Facility Change Request (FCR)78-317 was written to revise Be'chtel Drawing E-15,.MOV Data List, torque switch dial settings for DII 2733 and DH 2734 from 2.25 to 3.0.

DH 2733 is the suction isolation valve from the Borated Water Storage Tank (BWST) for DHR Pump 1-1 and DH 2734 is the suction iso-lation valve for DHR Pump 1-2.

This change was incorporated by Bechtel Company, the unit architect / engineer, in revision 5 of Drawing E-15, dated February 2, 1979.

REASON FOR THE FCR:

It was found that the previous torque switch dial setting of 2.25 was not adequate to properly seat the valves. Leak thru was occuring due to improper seating of the valves. Thr torque switch dial setting of 3.0 allows prcper seating and minimum leakage.

SAFETY EVALUATION:

Raising the recommended torque switch setting from 2.25 to 3.0 has been evaluated and was found to be acceptable. This setting provides satisfac-tory operation and is well below the maximum torque switch setting (set at 3.0 or 450 f t - lb versus maximum of 3.75 or 850 f t - lb). This change will not adversely affect the safety function of the decay heat removal system. Valves DH2733 and DH2734 are normally open valves. This is not an unreviewed safety question.

FACILITY CHANGE REQUEST COMPLETED DURING MAY,1979 FCR NO:

78-498 SYSTEM:. Emergency Core Cooling System (ECCS)

COMPONENT: High Pressure Injection (HPI) Pumps and Low Pressure Injection (LPI)

Pumps CHANGE,' TEST, OR EXPERIMENT; FCR 78-498 was written for the purpose of initiating a review to determine the proper interpretation of Subsection 6.3.3.2.3 of the Final Safety Analysis Report (FSAR). This subsection states that, "In the analysis of the loss of coolant accidents, the high pressure and low pressure injection was delayed 30 seconds after receipt of the actuation signal". The analysis was conducted by Bechtel, the unit architect / engineer, and Babcock & Wilcox, the nuclear steam Japply system vendor.

REASON FOR THE FCR:

The statement in FSAR Subsection 6.3.3.2.3 could be misinter-preted to imply that the LPI pump and HPI pumps would provide full rated flow within 30 seconds of receipt of the actuation signal. This is not a true statement for breaks smaller than the design basis LOCA as muount of flow will be determined by the Reactor Coolant System pressure.

For the design basis LOCA, the ECCS will pro-vide the full rated flow within 30 seconds of receipt of the actuation signal.

SAFETY EVALUATION:

A review of FSAR Subsection 6.3.3.2.3 and the Babcock & Wilcox safety evaluation as presented in B&W Topical Reportc BAW-10105 (ECCS Evaluation of Babcock & Wilcox's 177 FA Raised - Loop NSS) and BAW-10075 (Multinode Analysis of Samil Breaks for Babcock & Wilcox's 177 - Fuel - Assembly Nuclear Plants With Raised Loop Arrangement and Internal Vent Valves) indicates that the requirements for rated HP1 and LPI flow within 30 seconds is based on the NSSS design basis LOCA, which is an 8.55 f t.2 break in the reactor coolant pump discharge piping.

Any break of a smaller size requires less ECCS flow than the design basis event.

The correct interpretation of this FSAR subsection is that the system valves will be in their commanded (open) position in 30 seconds and the pumps (HPI and LPI) will be running at or above the flow corresponding to that assumed in Babcock &

Wilcox Topical Reports BAW-10105 and BAW-10075. This interpretation is consistent with the analysis and does not constitute an unreviewed saf ety question.

9 e 3/

o