ML19259C082

Monthly Operating Rept for May 1979

ML19259C082
Person / Time
Site:	Fort Calhoun
Issue date:	06/07/1979
From:	Hickle B OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19259C081	List: ML19259C080 ML19259C082
References
NUDOCS 7906120336
Download: ML19259C082 (9)

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AVERACE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun #1 DATE June 7, 1979 B. J. Hickle COMPLETED BY 402-536 kh13 TELEPilONE

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MONT11 MW-1070 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 453.0 h52.0 37 452.5 450.8 2

33 3

h53.3 h51.8 gg h53.7 451.5 4

20 h5h.0 h51.2 5

21 453.h hh9.6 6

22 h5].h 23 h50.2 7

8 h52.3 24 h51.2 9

h51.7 25 h51.h to h52.8 26 h50.6 11 h 53. h h50.6 27 h52.8 2g hh9 9 12 13 452.6 bb9 9 29 h52.9 hh8.9 14 30 453.3 4h9 9 15 3g h52.8 16 INSTRUCTIONS On this torrnat. list the average daily unit power level in MWe Net for each day in the reporting month. Compute io the nearest whole megawatt.

2285 028 19/77) 7 9 0 612 0 3 Sc,

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OPERATING DATA REPORT DOCKET NO. f 0-285 DATE June LJ79 COMPLETED BY B.

.I. Hickle TELEPllONE E02 '; 50-4413 OPERATING STATUS Notes

1. Unit Name:

Fort Calhoun Station Unit No. 1

2. Reporting Period:

May, 1979

3. Licensed Thermal Power (MWt):

th20

4. Nameplate Rating (Gross MWe):

502 457

5. Design E:ectrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe):

h81 457

7. Maximum Dependable Capacity INet MWe):

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

N/A

9. Power Level To'Which Restricted. If Any (Net MWe):

N/A

10. Reasons For Restrictions. If Any:

N/A This Month Yr.-to-Date Cumulatise

11. Ilours in Reporting Period Thb.0 3.623.0 h9,800.0

12. Number Of Hours Reactor Was Critical 7kh.0 3.61h.3 10.570.3

13. Reactor Reserve Shutdown flours 0.0 0.0 1,136.0

14. Ilours Generator On Line 744.0 3,602.3 38,668.h 0.0 0.0 0.0

15. Unit Resene Shutdown Hours

16. Gross Thermal Energy Generated (MWil) 1,0h5,546.3 h,956,223.8 h6,604,6h7.1

17. Gross Electrical Energy Generated (MWill 352,903.8 1,677,029.9 15,481,291.6

18. Net Electrical Energy Generated (MWil) 336,110.9 1,595,783.2 1h,617,188.0

19. Unit Senice Factor 100.O 00,h 77.6

20. Unit Availability Factor 100.0 99.4 77.6

21. Unit Capacity Factor IUsing MDC Net) 98.9 96.4 6h.8

22. Unit Capacity Factor (Using DER Net) 98.-9 96.4 64.2

23. Unit Forced Outage Rate 0.0 0.6 4.6

24. Shutdowns Scheduled Oser Nest 6 Months (Type. Date.and Duration of Each):

N/A

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

N/A

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

Fourcast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPER ATION 2285 029 (9/77)

e UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-285 UNIT NAME Fort Calhoun #1 DATE

.T" a a 7-1070 COMPLETED By n.

.T. u< ou REPORT MONDI May, 1979 TELEPilONE 40?-536 4L!3 IE is w

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} }Yl Licensee P,

ja, Cause & Corrective No.

Date g

3 Ej

,y5 Event s?

o-3 Action to b

9, Report a iE0 5O Prevent Recurrence H

$5 5

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Q (D

NONE GEHD B33 b

N N

CD 1

2 3

4 LJ1 F: Forced Reason:

Method:

Exhibit G Instructions S: Scheduled A Equipment Failure (Explain) l-Manual for Preparation of Data O

B Maintenance of Test 2-Manual Scram.

Entry Sheets for Licensee u

C-Refueling 3 Automatic Scram.

Esent Report (LER) File (NUREG-O D-Regulatory Restriction Other ( Explain)

OlM i l-:-Operator Training & Licene Examination F-Adnumst ratise 5

G-Operational Error (Explai.i)

Estubit I-Same Source (9/77) 11 Other ( Explain)

Pcfueling Information Ebrt Calhoun - Unit Ib.1 o

Deport for the nonth ending May 31, 1979 1.

Scheduled date for next refueling shutdown.

Jantiary 1,1980 2.

Scheduled date for restart following refueling.

March 1, 1980 3.

Will refueling or resumption of operation thereafter require a technical specification change or other Yes license amendment?

If answer is yes, what, in general, will a.

these be?

b.

If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Comittee to deter-mine whether any unreviewed safety questions are associated with the core reload, c.

If no such review has taken place, when is it scheduled?

Stretch Power Application 4.

Scheduled date(s) for submitting proposed licensing

. Site Related Information, action and support information.

July, 1979

.Non-Core Related Information 5.

Inportant licensing considerations associated with October, 1979 refueling, e.g., nca or different fuel design or

. Core Related Analysis and supplier, unreviewed design or performance analysis Tech. Spec. Changes, methods, significant changes in fuel design, new November, 1979 operating procedures.

6.

'1he number of fuel assenblies: a) in the core 133 assemblies b) in the spent fuel pool 157 c) spent fuel pool storage capacity h83 d) planned spent fuel pool storage capacity h83 7.

'1he projected date of the last refueling that can be discharged to the spent fuel pool assunting the present licensed capacity.

1985 Prepared by Date June 7, 1979

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2285 031

OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 May 1979 Monthly Operations Report I.

OPERATIONS SUlO!ARY Fort Calhoun Station Unit No. 1 operated at essentially a nominal 100% power during the month of May.

Simulator training at the Combustion Engineering Simulator for four hot license candidates and senior and operator licensed personnel continued through the month.

Nor=al operations and surveillance tests vere ec=pleted during the

=onth of May.

A.

PERFORMANCE CHARACTERISTICS LER Uumber Deficiency 79-010 During the performance of surveillance test ST-SI/CS-1, contain=ent spray pu=p SI-3B failed to start by control switch. The remaining two containment spray pu=ps were operable through-out this event. The unit was operating at approximately 100% steady state power.79-011 During nor=al power operation it was noted that Channel A DC Sequencer S1-1 had de-energized Matrix lights for engineered safeguards loads supplied from h80 volt bus 133A. The AC Sequencer for Channel A and both AC and DC Sequencers for Channel B vere operable through-out this event. The safeguards loads from the other plant h80 volt buses tied to Sequencer Sl-1 were also considered operable.79-012 During the performance of surveillance test ST-RPS-11 F.1 the matrix test failed for the BC = atrix. This test is required by Technical Specification Table 3.1 Ite 11.

The failure resulted in only a slight degradation as 5 of the 6 logic units were working properly when the fault resulted in a loss of one of four trip relays in one of the six asse=blies.79-014 As directed by IE Bulletin 79-01, ASCO solenoid valves were found tc it.ch sufficient environ-mental qualification data.

These solenoid valves function to provide or vent air to containment safety related valve operations. The plant operators have been instructed to fail instru=ent air te containment during post-LOCA conditions which potentially cause solenoid failure.

Failure of instrument air vill ensure that these safety related valves are =aintained in their safety position.

2285 032

Monthly Operations Report May 1979 Page Two A.

PERFORMANCE CHARACTERISTICS (Continued)

LER Number Deficiency 79-015 On May 11, 1979, the District was informed by its A/E (Gibbs & Hill) that seismic support appeared to be inadequate for the ec=ponent cooling water piping to centainment cooling and filtering unit VA-3A.

At 1630 on May 15, 1979, the preliminary indication was confirmed.

Also at this time, containment cooling and filtering unit VA-3A vas i==ediately declared inoperable in accordance with Technical Speci-fication 2.4 B.

CHANGES IN OPERATING METHODS None C.

RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS Surveillance tests as required by the Technical Specifications Sectica 3.0 and Appendix B, vere performed in..ccordance with the annual surveillance test schedule.

The following is a su==ary of the surveillance tests which results in Operations Incidents and are not reported elsewhere in the report:

Operations Incident Deficiency 0I-797 ST--ISI-CVCS-3 CH 4B discharge pressure high. Retested within specification.

Surveillance Test No.

Description ST-ISI-CVCS-3 (F.1)

Boric Acid Pu=p CH hB discharge pressure Che=ical and Volume was 100 psig which is in excess of 1.03 x P *r Control Syste= Pu=p Pu=p to be repaired.

Test D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT CCMMISG10N APPROVAL Procedure Descriotion SP-CE! 4-1 Spiked Radiological Sa=ples fer laboratory analysis comparison /cc=pleted per procedure remarks noted in co==ent section.

2285 033

Month Operations Report May 1979 Page Three D.

CHANGES, TESTS AND EXPERDiE:iTS CARRIED CUT WITHOUT COMMISSION APPROVAL (Centinued).

Procedure Description SP-RPS-5 Excore Detector Sy:=etric offset recalibration/

completed per procedure - no significant abnomal results.

DCR ThA-67 New Fuel Rack Modification (Boral Sheet)/

c;.=pleted as designed.

DCR 79-35/

DC0 79-17 Security Building Diesel Generator Roc =

lighting.

DCR 78-22 Turbine Building Jib Crane = Installed as designed.

DCR 77 h5 Upgrading Security Syste= - Completed as designed.

E.

PISULTS OF LEAK RATE T"STS None F.

CHANGES IN PLANT OPERATEIG STAFF Hone G.

TRADIDIG Training consisted of Simulator Training for the operators at Co=bustion Engineering.

Plant training included monitor team training, driver training, valve and pu=p packing, and procedure reviews.

H.

CHANGES, TESTS AND EXPERIMEtITS PIQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 None h/

/$&I!!Wouw Approved by Manager-Fort Calhoun Station 2285 034

Monthly Cperations Report May 1979 Page Four II.

MAIII'!!CICI (Significant Safety Related)

M.

O.

2 ate Oescription Corrective Action 400 h-9-79 AC Id.trix failure of Mercury Cycled relay.

Wetted Relay.

539 h-2h-7 9 DC Sequender for Bus 133A Relay Replaced relay.

Failure.

593 4-30-79 Turbine Building to Auxiliary Readjusted door.

Building fire door closer isn't verking.

380 h h-79 DG-2 Recirculating pu=p has Replaced faulty coupling.

568 h-30-79 Doer 1007-16 vill not close Adjusted properly.

162 5 h-79 Energency Diesel #2 radiator Repaired core and replaced core leaks.

gasket.

635 5-5-79 Emergency Diesel #2 safety valve Repair and cc=pleted, on #2 air ec= pressor.

6h5 5-7-79 Fire Protecticn - open floor Installed thresded pipe caps penetration in Roo 81 as needed.

637 5 8-79 Energency Diesel 12 - Relief Replace relief valve, valve en 11 air conpressor leaking through.

632 5 h-79 Energency Diesel #2 - crack in Repair and welded crack.

radiator vr.ter divider.

46 79 Heater Drain Pu=p (7W-5C) - install Cc=pleted throttle bushing bleed off line to j

ensure proper lubrication.

733 5-17-79 AC-pipe supports - install U bolts Cc=plet ed per procedure.

723 5-16-79 A-046 Load New Fuel for shipment.

Cc=pleted 65h 5-8-79 RPs Matrix relay BC 4 failure.

Removed and cleaned contacts.

723 5-16-79 A0h6 New Fuel Bundle load for Completed shipnent.

2285 035

Menthly Cperstions Report May 1979 Fage Five II.

lG::T2DE!CI (Significant Safety Related)

M. O. #

Date Description Corrective Acticn 733 5-17-19 Replace U-bolts on CCW piping to Uebolts installed per detailed containnent air coolers that were work instructions, left off during construction.

60h 5-1-79 Retest of CH h3.

Retested per ST-ISI.CVCS-3 (F.1).

Discharge press. = 98 psig, within specification.

777 5-25-79 Seismic Support FWS-83 Install new U-bolts.

2285 036

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NUCLEAR ENERGY 1,IABILITY INSURANCE MUTUAL ATOMIC ENERG3 LIABILITY UNDERWRITERS 1.

Amendment of Advance Premium Endorsement 2.

Standard Premium and Reserve Premium Endorsement 3.

Additional Premium Due 1.

Advance Premium It is agreed that the Amended Advance Premium due the companies for the calendar year 1979 is $96,870.95 2

Standard Premium and Reserve Premium Subject to the provisions of the Industry Credit Rating Plan, it is agreed that the Standard Premium and Reserve Premium for the calendar year designated above are:

Standard Premium

$96,870.95 Reserve Premium

$72,946.92 3

Additional Premium

$4,726.93 Effective Date of To form a part this endo rsement Janua ry 1,1979 of Policy No.

MF-57 Issued to Mrginia Electric and Power Company Date of Issue May 10,1979 For the Subscribing Companies MUTUAL ATOMIC ENERGY LIABILITY UNDERWRITERS By a,

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Countersigned by m m cr c a E g $ f Authorized Representative hb E% ELM" gA(

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s TENNESSEE VALLEY AUTHORITY CH ATTANOOGA TENNESSEE 37401 716 Edney Build!us June 6, 1979 Mr..fannes P. O'Railly, Director U.S. k% clear Regulatory Cocaissi m Office of Ins,estion and Ep'-

mt Region II 101 Marietta Street. Suite 3100 Atlanta, Georgia 30303

Dear Mr. O'Railly:

TENNESSEE VALLEY AUTHORITY - BROWNS TERAT NUCLEAR PLANT UNIT 2 - DOCIET ro. 50-260 - FACILITY OPERATING LICENSE DPR REPORTA3LE OCCURRENCE REPORT BFRO-50-260/7911 The enclosed report provides dormila concerning the first rod, 26-27, in group 3 which was inadvertently withdrawn three notches while pulling control rods to achieve initial criticality for cycle 3.

This report is submitted in accordance with 3rovna Ferry unit 2 Technical Specification 6.7.2.a.4.

Very truly yours, TEICIESSEZ VALLZT AUTHORITT H. S. Fox Director of Power Production Enclosure (3) cc (Enclosure):

Director (3)

Offica of Management Information and Program Control U.S. Hucisar Regulatory Commission Washington, D.C.

20555 Director (40)

Office of Inspection and Enforceman" U.S. Nuclear Regulatory Commission Washington, D.C.

20555

1. r...uiv.a. nic ras,ector, >-n.1."'

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a 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75 RgPORT oATE 80 EVENT DESCRIPTION AND PROB ABLE CON 5sOUENCES h While pulling control rods to achieve initial criticality for cycle 3, the first o

2 o a i rod, 26-27, in group 3 was inadvertent'.y withdrawn three notches. The rod moved from i 10141 Lposition 02 to position 03 rather *:.n e.3m position 02 to 04 as prescribed in the i

i l rod move sheets. The reacta. scrammed on hi-hi IRM flux. Based on evaluation of l

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o s l process neutron monitor ng instrumentation no safety limit was exceeded but is 1

l reported in accordance with T. S. 6.7 '.ab No hazard existed to the health and

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i safety of the public. Previous occurrence: 259/7901.

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A 31 40 41 42 43 44 47 CAusE oEscRiPTico ANo CCRRECTtvE ACTIONS l The withdrawal of the rod from position 02 to position 08 is unexplained. Functional l i

o testing of the rod after the event revealed that the rod would not travel three i

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positions when given a notch withdraw signal. Prior to restart, a review of the I

y l event was conducted and no safety limits were exceeded. A subsequent rod pull, l

3 l using the same sequence, resulted in a period of approximately 43 seconds.

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NAME OF PREPARER PHONE

e t-l Tennessee Valley Authority Form BF-17 Browns Ferry Nuclear Plant BF 15.2 1/10/79 LER SUPPLEMENTAL INFORMATION BFRO 260 / 7911 Technical Specification Involved 6.7.2.a.4 Reported Under Technical Specification 6.7.2.a.4 Date of Occurrence _5/26/79 Time of Occurrence 1928 Unit 2

Identification and Description of Occurrence:

BFNP unit 2 scrammed at 1928 on May 26, 1979, while pulling control rod 2,6-27, y

the first rod in RWM group 3, during an approach to critical. The cause of the scram was high neutron flux on IRM's C, D, and F.

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The ceactor period was calculated to be 5.5 seconds.

Conditions ' Prior to Occurrence:

I-Normal startup procedure following the spring 1979 refueling outage. This was the first approach to critical.

Action specified in the Technical Specification Surveillance Requirements met due to inocerable equipment. Describe.

N/A Apparent Cause of Occurrence:

Movement of control rod 26-27 three positions.

Analysis of Occurrence:

A reactivity insertion of approximately 0.287 percent delta K/K resulted when control rod 26-27 was withdrawn from position 02 to position 03. The resulting reactor period by the SRM recorder 74as aFout,one second. It was calculated to be 5.5 sec.onds.

Corrective Action: Pr'ior to initial criticality folloid.ng a refueling outage, any control rod that has a single-notch worth which is capable of producing a period of equal to or less than 60 seconds, or any control rod that has a double-notch worth capable of producing s period of equal to or less than 30 seconds, will be with-drawn'on a notch basis. The nuclear engineer will.be present in the control room during the period of time from initial rod pull until criticality is achieved.

During this period of time, he will confirm acceptable core behavior.

Failure Data:

N/A l

• Retention :

eriod - Lifetime; Responsibility - Administrative Supervisor D

2285 040

• Revision:

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