ML20205G986

Monthly Operating Rept for Jul 1986

ML20205G986
Person / Time
Site:	Fort Calhoun
Issue date:	07/31/1986
From:	Matthews T OMAHA PUBLIC POWER DISTRICT
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ML20205G951	List: ML20205G947 ML20205G986
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NUDOCS 8608190669
Download: ML20205G986 (17)

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AVERAGE DAILY UNIT POWER LEVEL 50-285 DOCKET NO.

Fort Calhoun Station UNIT August 11, 1986 DATE COMPLETED BY T P Matthewss 402-536-4733 TELEPflONE July, 1986 MONTli DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 464.4 17 466.7 466.6 94 7 18 2

465.9 0.0 39 3

465.4 76.8 20 4

467.4 263.7-21 5

6 asnA 22 46R s 468.6 7

455.4 23 467.8 457.6 24 8

460.6 467.1 25 9

to 469 3 26 428.7 463.6 470.2 27 468.9 464.4 12 23 469.6 464.6 13 7,

469.0 465.0 34 39 466.8 467.5 3,

15 16 467.1 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.

(9/77)

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OPERATING DATA REPORT DOCKET NO.

50-285 DATE Ana. 11. 1986 COMPLETED BY T. P. Matthews TELEPilONE 402-536 4733 OPERATING STATUS Notes Fort Calhoun Station

1. Unit Name:

July, 1986

2. Reporting Period:

150

3. Licensed Thermal Power (MWt):

4. Nameplate Rating (Gross MWe):

478

5. Design Electrical Rating (Net MWe):

502

6. Maximum Dependable Capacity (Gross MWe):

478

7. Maximum Dependable Capacity (Net MWe):

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

N/A N/A

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

None

10. Reasons For Restrictions,if Any:

This Month Yr..to-Date Cumulative 744.0 5,087.0 112,633.0 i1. Ilours in Reporting Period

12. Number Of flours Reactor Was Critical 711.2 4,858.8 86,605.0 0.0 0.0 1,309.5

13. Reactor Reserve Shutdown flours 703.1 4,674.8 85,797.7

14. Ilours Generator On Line 0.0 0.0 0.0

15. Unit Reserve Shutdown llours

16. Gross Thermal Energy Generated (MWil) 1 00% 26R 7 6,206,809.7 109,957,854.3

17. Gross Electrical Energy Generated (MWil) 327 442 1, 2.052,799.2 36.037.368.2

18. Net Electrical Energy Generated (MWil) 311,118.2 1,955,740.2 34.433,633,6 94.5 91.9 76.2

19. Unit Service Factor 94.5 91.9 76.2

20. Unit Availability Factor 87.5 80.4 66.2

21. Unit Capacity Factor (Using MDC Net) 87.5 80.4 64.2

22. Unit Capacity Factor (Using DER Net) 4.5 0.7 3.4

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Over Next 6 Months (Type.Date.and Duration of Each):

Nono

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

N/A

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

Forecast Achieved INITIAL CRITICALITY tyA INITIA L ELECTRICITY COMM ERCIAL OPER ATION (9/77) s J

UNITSHUTDOWNS AND POWER REDUCTIONS

' Fort houn Station N

E DATE Auaust 11. 1986 COMPLETED BY I.

P. Matthews REPORT MONTH July' 1986 TELEPHONE 402-536-4733 e.

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Cause & Corrective j

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Date g

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Action to y

35g Report e mO Prevent Recurrence H

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86-03 860702 F

39.3 A

3 LER-8601 EF INUT On July 2, 1986, at 0534, the reactor and turbine-generator were auto-matically tripped on low steam gener-ator level af ter the failure of a safety related instrument inverter.

The unit returned to service July 3, at 2052. See LER-86-01 for details of the corrective actions taken.

86-04 060704 F

1.6 H

4 N/A xx xxxx On July 4, at 0138, the Turbine-gen-erator was taken off-line to fix an EHC leak. The unit returned to service the same day at 0316.

I 2

3 4

m F: Forced Reason:

Method:

Exhibit G - Inst ructions S: Schedufed A-Equipment Failure (Explain)

I-Manual for Preparation of Data B Maintenance of Test 2-Manual Scram.

Entry Sheets for Licensee C. Refueling 3-Automatic Scram.

Event Report (LER) File (NUREG-D-Regularory Restriction 4-Other (Explain) 0161)

E. Operator Training & License Examination F Administrative 5

G-Operational Error (Explain)

Exhibit I Same Source (9/77)

Il-Other (E xplainI L

Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending July, 1986 1.

Scheduled date for next refueling shutdown.

March, 1987

_ 2.

Scheduled date for restart following refueling.

May, 1987 3.

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

Yes a.

If answer is yes, what, in general, will these be?

Fuel supplier change from ENC to CE and possible Technical Specification change.

b.

If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee 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?

4.

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

February. 1987 5.

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.

New fuel supplier (Combustion Engineering) but the new fuel will be essentially the same as that supplied by Combustion Engineering for Cycle 5.

6.

The number of fuel assemblies: a) in the core 133 assemblies b) in the spent fuel pool 349 c) spent fuel pool storage capacity 729 d) planned spent fuel pool May be increased storage capacity via fuel nin consolidation 7.

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

1996 Prepared by W = Wffft:--_

Date Auaust 1. 1986 J

OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No.1 July, 1986 Monthly Operations Report I.

OPERATIONS

SUMMARY

Fort Calhoun Station tripped off line on July 2,1986, for the first time in almost two years. The trip was caused by an instrument inverter failure. The reactor was critical the same day, but was shutdown again to replace a faulty CEDM package. On July 3, the plant was back on line. On July 4, the generator was taken off line for a short period to fix an EHC leak. The plant attained 100% power again on July 6.

Power was reduced temporarily to 80% on July 26 to repack a circulating water pump.

The NRC was onsite one week for an audit of the Emergency Response Facility.

Main feedwater pump FW-4B had its impeller installed and was placed back in service.

Installation of the water plant sampling upgrade modification was started in July. One waste concentrate cask was filled and three were shipped off site in July.

Training continued for licensed operator requalification, non-licensed operator refresher, chemistry / health physics refresher and general employee training.

Initial training took place for C/RP shift technicians and equipment operator nuclear-turbine building.

Training was given to electricians on circuit breaker operation and to mechanical maintenance personnel on flood protection.

Training was also given to instrument and control technicians on level instrumentation.

One PORV challenge did occur due to the above mentioned reactor trip.

Both PORV's reseated properly following actuation. Visual inspections of discharge piping and piping supports were conducted with no abnormalities found. Reference memo FC-886-86 (revised 7-9-86) for further information concerning the July 2, 1986, plant outage.

No Pressurizer Safety Valve challenges or failures occurred during the month.

A.

PERFORMANCE CHARACTERISTICS None B.

CHANGES IN OPERATING METHODS None C.

RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS None 4

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r-Monthly Operations Report July, 1986 Page Two

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D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Procedure Description SP-FAUD-1 Fuel Assembly Uplift Condition Detection.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since it only involved the evaluation of data from a surveillance test to verify that a fuel assembly uplift condition did not exist.

SP-FAUD-1 Fuel Assembly Uplift Condition Detection.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since it only involved the evaluation of data from a surveillance test to verify that a fuel assembly uplift condition did not exist.

SP-HJTC-1 Electrical Preoperational Tests.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since this procedure merely outlines and documents functional checkout and testing of a heated junction thermocouple assembly.

SP-HJTC-1 Electrical Preoperational Tests.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since this procedure merely outlines and documents functional checkout and testing of a heated junction thermocouple assembly.

SP-CEA-4 CEA Inspections This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because the procedure only provided for inspections of control element assemblies and analyzing of the data.

SP-CEA-4 CEA Inspections This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because the procedure only provided for inspections of control element assemblies and analyzing of the data.

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Monthly Operations Report July, 1986 Page Three l

D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

SP-SFS-2 Receipt of Spent Fuel in NLI 1/2 Cask This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because it only addresses receipt of irradiated fuel.

Procedure requires use of a safe load path, prohibits transporting the cask over irradiated fuel and addresses use of charccal filters. Drop of an irradiated fuel assembly is addressed in Section 14.18 of the USAR.

System Acceptance Committee Packages for July, 1986:

Packaoe Description / Analysis DCR 76-26 HE-2 Single Fail Proof Trolley.

This modification provided for the installation and testing of the retro-fit trolley system and existing overhead crane bridge modifications for the refueling area crane.

This modification does not have an adverse effect on the safety analysis.

DCR 77-20 Waste Concentrate Pump Local Control WD-39A/B.

This modification provided for local control switches at each concentrate pump, WD-39A/B. This modification does not have an adverse effect on the safety analysis.

DCR 77-97 Seal Water for Raw Water and Circulating Water Pumps.

This modification replaced corroded potable water and service water lines with new stainless steel pipe. Two spare lines were also run for future use. Two booster pumps were installed in the SW line to increase seal water line pressure. This modification does not have an adverse effect on the safety analysis, f

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Monthly Operations Report July, 1986 Page Four D.

CHANGES, TESTS AND EXPERIMENTS CARRIED Odf WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July,1986:

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Packaae Description / Analysis

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EEAR FC-78-14 Potable Water Line to > Intake Structure.

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This modification'repla ed corroded potable water and service water lines gith ndmstainless steel pipe.

Two spare lines werealso, run for future use.

Two booster pumps were installed in the SW line to increase seal water line pressure.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-78-17 Seal Water Modification to Raw Water Pumps.

3 This modification replaced corroded potable water and service water lines,with new stainless steel pipe. Two spare lines were also run for future use. Two booster pumps were installed in the SW line to increase seal water line pressure. This modification does not have an adverse effect on the safety analysis.

RC-3CLubeOil'CoolerPipeSupport.}

EEAR FC-78-063 This modification provided for the installation' of two snubbers to RC-3C's lube oil cooler piping.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-79-036 Nuclear / Fire Alarms Addition to Maintent.nce ' Shop and Security Building.c ;

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This modification provided for the installation of nuclear / fire alarns in the maintenance shop (see FC-85-027 for security building modification). This modification does not have an adverse effect on the safety analysis.

EEAR FC-79-216 Maintenance Shop Sprinkler System Alarm.

3 This modification provided for the installation of a flow alarm on the maintenance shop sprinkler system.

This modification does not have an adverse effect on the safety ' analysis.

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Monthly Operations Report July, 1986 Page Five D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July, 1986:

Packaae Descriotion/ Analysis EEAR FC-80-043 RC Pump Insulation.

This modification provided for the replacement of the Kaylo block insulatica with blanket insulation.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-80-106 Structural Modification to Observation Room.

This modification structurally modified the observation room to meet NRC IE Bulletin 79-16 requirements. This modification does not have an adverse effect on the safety analysis.

EEAR FC-80-128 Plant Site Erosion Control.

This modification provided for improved drainage on the plant site to minimize erosion of areas surrounding the plant access road. This modification does not have an adverse effect on the safety analysis.

EEAR FC-80-144 Sprinkler System Flushing Valves.

This modification provided for the installation of sprinkler flushing valves on branch lines in the turbine building sprinkler system. This modification does not have an adverse effect on the safety analysis.

EEAR FC-81-016 Communication System.

This modification provided for the installation of the new ROLM phone system in the plant. This modification does not have an adverse effect on the safety analysis.

EEAR FC-81-074 Communications Equipment Room.

This modification provided for the area expansion of the communications room to allow for the installa-tion of new equipment; specifically, telephone switching equipment and a 48V battery bank. This modification does not have an adverse effect on the safety analysis.

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Monthly Operations Report July, 1986 Page Six D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July, 1986:

Packaae Descriotion/ Analysis EEAR FC-81-087 Emergency Response Facilities Computer System.

This modification provided for the installation of the ERF computer system. This modification does not have an adverse effect on the safety analysis.

EEAR FC-81-104 Copier Electrical Service.

This modification provided for the installation of a separate power receptacle for the storeroom copy machine. This modification does not have an adverse effect on the safety analysis.

EEAR FC-81-135A Quality Assurance File Room Upgrade.

^ This modification provided for the upgrade of the QA file room to meet ANSI N45.2.9-1979 standards-(installation of Halon 1301 suppression system).

This modification does not have an adverse effect on the safety analysis, i

EEAR FC-81-135B Quality Assurance File Room Fire Detection Panel Annunciation.

This modification provided for the installation of wiring and equipment to provide remove-annunciation of fire detection panel AI-209. Also included testing of the annunciation system. This modification does not have an adverse eff,cet' on the safety analysis.

EEAR FC-81-147 FH-12 Cable Slack Bypass.

.4 7

This modification provided for the installation of a switch to bypass the cable slack interlock. This enables the operator to lower an empty hook withoJt encountering an interlock. This modification does not have an adverse effect on the safety analysis.

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Monthly Operations Report July, 15'S5 Page Seven D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July, 1986:

Packaae Description / Analysis EEAR FC-81-178 Reactor Coolant Gas Vent System.

This modification provided for the installation of a blank flange on the end of the reactor coolant gas vent system atmospheric vent to provide isolation in case of valve leakage and to provide a barrier for hydrotesting. This modification does not have an adverse effect on the safety analysis.

4 EEAR FC-82-030 TCV-202 Leakage.

This modification provided for the addition of isolation valve (CH-450) to TCV-202 leakoff line.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-82-044 Outage Scheduling Room.

This modification provided for the installation of an outage scheduling room in the service building.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-010 Connecting Radiation Monitor Inputs to ERF Computer System.

This modification connected radiation monitors to the ERF computer system. This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-040 Sewage Treatment Lagoon Well.

This modification provided for the installation of a well to maintain a required level in the sewage treatment lagoons. This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-049 Conversion of HCV-884A Seat Material.

This modification provided for the replacement of the Disc material in valve HCV-884A with Viton. The original material caused the valve to leak. This modification does not have an adverse effect on the safety analysis.

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Monthly Operations Report July, 1986 Page Eight D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July, 1986:

Packaoe Description / Analysis EEAR FC-83-109 Transfer of P250 Points to ERF Computer.

This modification transferred points from P250 to ERF Computer. This modification does not have an adverse effect on the safety analysis.

I EEAR FC-83-116 Vibrations and Loose Parts Monitoring System.

This modification provided for the installation of components for the vibrations and loose parts monitoring system. This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-129 DG-2 Speed Sensing Circuitry Power Supply.

This modification installed potentiometer in place of failed thermistor as a temporary fix until new thermistor could be acquired. This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-152 DG-2 Speed Sensing Circuitry Power Supply.

This modification provided for changing the power supply and DC feed was provided for the specd sensing circuit. This modification does not have an adverse effect on the safety analysis.

EEAR FC-83-168 Replacement of AS-618.

This modification provided for the replacement of AS-618 (strainer and trap bypass valve for auxiliary boiler lower drum) with a 3/4" Vogt socket weld gate valve. This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-001

" Spent Resin Tank Level Channel 639.

This modification provided for the replacement of old transmitter ST-10 with a new ST-25 Milltronics transmitter and added two sigma lumigraphs. This modification does not have an adverse effect on the safety analysis.

Monthly Operations Report-July,1986 Page Nine D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT CCMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July, 1986:

Packaae Descriotion/ Analysis EEAR FC-84-028 Annunciator Update-SI; PAS-2902, 2922, 2942 & 2962.

This modification replaced pressure switches PAS-2902, 2922, 2942 and 2962 with switches with smaller dead banks on SI tanks 6A, 6B, 6C and 6D.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-041 Loops 101X/Y Computer Interface.

This modification provided for the installation of isolated transmitters on instrument loops 101X and 10lY to solve the problem of noise spikes caused by computer scanning of the loops. This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-077 Replace HCV-335 Solenoid and Limit Switch.

This modification provided foi the replacement of solenoids and limit switches with CQE grade equipment for HCV-335. This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-084 DC Grounds on HCV-5068, 2907, 2908 and 2918.

This modification placed a varistor across the solenoids to prevent grounds during cycling.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-90 PORV Reset Demand.

This modification will eliminate PORV pre-trip alarms when PPLS is unblocked.

Reactor coolant overpressure protection - channels 105/123 and 113/115. This modification does not have an adverse effect on the safety analysis.

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Monthly Operations Report July, 1986 Page Ten D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July,1986:

Packaae Description / Analysis EEAR FC-84-096 Replacement of Safety Related HFA Relays.

This modification provided for the replacement of approximately 160 HFA relays in response to NRC IE Bulletin 84-02. This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-113 HCV-545 Accumulator.

This modification disconnected air accumulator from HCV-545. This modification does r.ot have an adverse effect on the safety analysis.

EEAR FC-84-140 Delta T Power Process Loops.

This modification provided for the installation of new RTD's in the RPS delta T power process loops.

This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-144 Upgrade of YCV-1045A/B Solenoid.

This modification provided for the replacement of ASCO solenoid valves with Class IE qualified ASCO valves for YCV-1045A and YCV-1045B. This modification does not have an adverse effect on the safety analysis.

EEAR FC-84-204 DC Grounds on HCV-5008 and HCV-5078.

This modification provided for the placement of varistors across solenoids to eliminate spurious trips. This modification does not have an adverse effect on the safety analysis.

EEAR FC-85-006 Annubar.

This modification provided for the movement of the annubar from Room 59 to Room 60 for operator convenience. This modification does not have an adverse effect on the safety analysis.

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Monthly Operations Report July, 1986 Page Eleven D.

CHANGES, TESTS AND EXPERIMENTS CARRIED Odf WITHOUT COMMISSION APPROVAL (continued)

System Acceptance Committee Packages for July,1986:

EEAR FC-85-165 TSC Communications.

This modification provided for the installation of two new Gai-tronics in the TSC and installed a paging system exclusively for the TSC. This modification does not have an adverse effect on the

]

safety analysis.

I EEAR FC-86-19A Site Security Building Expansion - Fence Relocation.

This modification moved the security fence i

temporarily to accommodate expansion of the security building. This modification does not have an adverse effect on the safety analysis.

EEAR FC-86-42 Revised Power Supply for AI-50.

This modification rearranged the power feed.for AI-50 from inverter A to inverter 2.

This modification does not have an adverse effect on the safety analysis.

E.

RESULTS OF LEAK RATE TESTS The biannual Containment Pal Door leak rate test was completed during the month of July per ST-CONT-2. The total leakage for the PAL doors was found and left at 0 SCCM.

Attheendofthe1985refuelingoutage,thekn'Ldoorsleakagewas found and left at 1000 SCCH. As of June,1986, after the biannual purge valves test, the total "B" and "C" leak rate was 5,443 SCCM. The new lcwer leak rate on the PAL door drops the total accumulative "as left" leak rate to 4,443 SCCM.

The new "B" and "C" total leak rate of 4,443 SCCM is well below the allowed leakage of.6 La (62,951 SCCM) as specified in 10 CFR 50 Appendix J.

The next scheduled test that will affect the "B" and "C" leak rate total is the biannual purge valves test.

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Monthly Operations Report July, 1986 Page Twelve i

4 F.

CHANGES IN PLANT OPERATING STAFF None.

G.

TRAINING i

Training continued for licensed operator requalification, non-licensed operator refresher, chemistry / health physics refresher and general employee training.

Initial training took place for C/RP shift technicians and equipment operator nuclear-turbine building. Training was given to electricians on circuit breaker operation and to mechanical maintenance personnel on flood protection. Training was also given to instrument and control technicians on level instrumentation, i

l H.

CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 i

Amendment No.

Description 98 The amendment revises Sections 2 and 3 of the Technical Specifications to incorporate operability i

and surveillance requirements for new fire suppression equipment in the compressor room.

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II. MAINTENANCE (Significant Safety Related) 4 None

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M W. Gary bates-i Manager Fort Calhoun Station i

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