ML20090H488

Monthly Operating Rept for June 1984

ML20090H488
Person / Time
Site:	Fort Calhoun
Issue date:	06/30/1984
From:	Andrews R, Matthews T OMAHA PUBLIC POWER DISTRICT
To:	Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
LIC-84-213, NUDOCS 8407260318
Download: ML20090H488 (11)

Text

_

o AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Static DATE July 6,1984 COMPLETED BY T. P. Matthews TELEPilONE (402) 536-4733 MONTH June, 1984 DAY AVER AGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 3 0.0 37 0.0 2 0.0 18 0.0 3 0.0 19 0.0 4 0.0 20 0.0 5 0.0 21 0.0 6 0.0 22 0.0 7 0.0 23 0.0 8 0.0 24 0.0 9 0.0 25 0.0 10 _

0.0 26 0.0 gg- 0.0 27 0.0 12' O.0 28 0.0 g, 0.0 29 0.0 g4 0.0 30 0.0 15 0.0 3, 16 0.0 INSTRUCTIONS On this fortnat. list the average daily unit power levelin MWe Net for each day in the reporting month.Cornpute to the nearest whole megawait.

(W771 8407260318 840630 PDR ADOCK 05000285 R PDR

I '

0' Y

f OPERATING DATA REPORT DOCKET NO. 50-285 DATE _ July 6,1984 COMPLETED BY _T. P. Matthews TELEPilONE (402) 536-4733 OPERATING STATUS Ebrt Calhoun Station N"te$

1. Unit Name:

2. Reporting Period: June, 1984

3. Licensed Thermal Power IMW ):

1500

4. Nameplate Rating (Gross MWe):

501

5. Design Electrical Rating (Net MWe): 470 501

6. Masimum Dependable Capacity (Gross MWe):

7. Maximum Dependable Capacity (Net MWe):

478

8. If Changes Occur in Capacity Ratings (Items Nu:: ber 3 Through 7) Since Last Report.Give Reawns:

6 and 7 were restored to their previous values to reflect replacement of the first stage blading in the high pressure turbine.

9. Power Lesel To which Restricted.If An iNet MWe): N/A

10. Reasons For Restrictions.lf Any: ne This Month Yr..to Date Cumulative

11. Hours in Reporting Period 720.0 4,367.0 94,369.0

12. Number Of Hours Reactor Was Critica! 0.0 1,490.2 71,384.1

13. Reactor Reserve Shutdown flours 0.0 0.0 1,309.0

14. Hours Generator On Line 0.0 1,489.5 70,892.0

' 15. Unit Reserve Shutdown Hours . 0.0 0.0 0.0

16. Gross Thermal Energy Generated (MWill 0.0 2,152,796.9 88,912,510.6 17 Grow Electrical Energy Generated (MWHI 0.0 690,258.0 _29,007,827.0

18. Net Electrical Energy Generated (MWH1 0.0 656,536.5 27,736,405.2

19. Unit Service Factor 0.0 34.1 75.1

20. Unit Asailability Factor 0.0 34.1 75.1

21. Unit Capacity Factor (Using MDC Net) 0.0 31.5 64.1

22. Unit Capacity Factor (Using DER Nel1 0.0 31.5 61.8

23. Unit Forced Outage Rate 0.0 0.0 3.5

24. Shutdowns Scheduled Oser Nest 6 Months ITy pe, Date,and Duration of Each t

25. If Shut ihmn At End Of Report Period Estimated Date of Startup: July _.10, 1984

26. Units in Test Status iPrior to Commeseial Operationi: N/A Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICifY COMMERCIAL 6PER ATION 19/77#

L

M DOCKET NO. 50-285 UNIT SHUTDOWNS AND POWER REDUC 110NS UNIT NAME Fort Calhoun Station DATE EV 6,1984 COMPLETED BY T. P. Matthews REPORT MONTH ~Jime, 1984 TELEPHONE (402) 536-4733

. r

= c. 3 .a 5a Cause & Corrective 2 _: 2 E4 Eicensee c.

E

• c .,

No. Date g 3g g jg5 Event s? E.3 Action io

$. 3 5 ji2j Report

• NC U Prevent Recurrence 2

84-01 840303 S 2877 c 4 N/A XX XXXXX 1984 refueling outage eminenced March 3, 1984.

I 2 3 4 F: Forced Reason: Method: Exhibit G - Instructions S: Scheduled A-Fqmpment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram- Entry Sheets for Licensee C Refueling 3 Automatic Scram. Event Report iLE RI File (NUREG-D-Regulatory Restriction 4 Other (Exp aini 01611 F-Operator Training & License Esamination F-Administrativ 5 CrOperational Error (Explain t Eshibit I - Same Source 19/77) Il-Other (E splain)

9 Refueling Information Ebrt Calhoun - Unit No.1 Report for the month ending June 1984 .

1. Scheduled date for next refueling shutdown. September 1985

2. Scheduled date for restart following refueling. November 1985

3. Will refueling or resunption of operation thereafter

> require a technical specification change or other

-license amendment? Yes

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

b. If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Comnittee 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. August 1985

5. Inportant 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.

6. We number of fuel assenblies: a) in the core 133 assm blics b) in the spent fuel pool 305 c) spent fuel pool _

storage capacity 729 d) planned spent fuel pool * "

storage capacity

• fiay be increased via fuel pin consolidation

7. %e projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity. 1996 Prepared by _

et -pl/) Date July 2, 1984

^

V f

• e OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No.1 June, 1984 lionthly Operations Report

.I. OPERATIONS

SUMMARY

Fort Calhoun Station has been shutdown through June for testing and repair of the steam generators. Plant startup began on June 28 and the plant should go enline in July.

Three operators sat for NRC reactor operator license examinations in June.

The Missouri River reached a high level of 1002.4' MSL on June 27 flooding the chemical waste lagoons and part of the parking lot.

No safety valve or PORY challenges occurred.

A. PERFORMANCE CHARACTERISTICS Hone B. CHANGES IN OPERATING METHODS Hone C. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS Hone D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Procedure Description SP-RC-2-1 Plugging Steam Generator Tubes (RC-28).

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provides for installation of two tube plugs in "B" steam i generator. Tubes were plugged in accordance with Combustion Engineering guidelines.

Monthly Operations Report

-June, 1984 Page Two D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL

{ Continued)

Procedure Description

'SP-RC-2-1 Plugging Steam Generator Tubes (RC-28).

This procedaro did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provides for installation of three tube plugs in "B" steam generator. Tubes were plugged in accordance with Combustion Engineering guidelines.

SP-RC-2-1 Plugging Steam Generator Tubes (RC-2B).

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provides for installation of four tube plugs in "B" steam generator. Tubes were plugged in accordance with Combustion Engineering guidelines.

SP-VA-80 Hydrogen Purge System Test.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provides for running of blowers and measurement of differential pressure. No technical specifications govern the use of this equipment.

SP-MS-4 Testing of Steam Dump Valves TCV-909-1, 909-2, 909-3 and 909-4.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because the steam dump valves are not considered to be safety related and even though they may help to mitigate the consequences of specific accidents, they are not credited for helping to mitigate the consequences of any accident previously . evaluated in the USAR.

i-L

Monthly Operations R2 port June, 1984 Page Three System Acceptance Committee Packages for June,1984:

Package Description / Analysis EEAR FC-82-91 Steam Generator Blowdown lionitor Relocation.

This modification provided for sufficient shielding of the steam generator blowdown monitors. Samplers with six -inch thick lead shield were installed because of high background radiation within the sampling room 60.

Therefore, the installation of the new samplers will improve the function of the monitors. This modification has no adverse effect on the safety analysis.

EEAR FC-79-66 Qualification of Backup Containment Instrumentation for Post LOCA.

This modification provided for the relocation of existing transmitters and the one-for-one functional replacement of RTD's. Channel redundancy, equipment separation and power supply independence are consistent with the existing systems. Measurement sensitivity is equivalent to the existing equipment. This modification has no adverse effect on the safety analysis.

EEAR FC-82-84 Capillary Tube Replacement on RCDT.

This modification provided for the replacement of a capillary tube on RCDT. This modification has no adverse effect on the safety analysis.

EEAR FC-83-50 Turbine Drain Valve Controls.

This modification provided for the installation of a drain valve to operate automatically upon turbir.e trip and to help prevent water induction in the llP section of the turbine. This modification has no adverse effect on the safety analysis.

EEAR FC-93-56 Hydrogen Purge Valves VA-289, VA-280.

This modification provided for the replacement of existing valves with valves designed for a better seal.

This modification has no adverse effect on the safety analysis.

. . .~

Monthly Operations Report June, 1984 Page Four D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

System Acceptance Committee Packages for June,1984: (Continued)

Package Description / Analysis DCR 748-38 Pressurizer Channel 101Y Output Oscillation.

This modification replaced transmitters on pressurizer level channels. This modification has no adverse effect on the safety analysis.

DCR 758-16 PT-235 Vibration Damage.

This modification improved the reliability of the system by moving the charging pump pressure transmitter and protecting it from vibration. This modification has no adverse effect on the safety analysis.

DCR 76-49 Flow Meters for Radiation Monitors.

This modification replaced rotameters with flow meters.

This modification has no adverse effect on the safety analysis.

DCR 77-39~ CEDM Cooling Ducts.

This modification provided for the relocation of a joint in the ductwork to a more accessible locatior,.

This modification has no adverse effect on the safety analysis.

DCR 78-7 Generator Core Monitoring.

This modification added a means to detect early stages of generator core overheating. This modification has no adverse effect on the safety analysis.

~

EEAR FC-78-29 RPS Noise Spikes.

This modification improved the reliability of the RPS

^^

system by eliminating spurious trips caused by the operation of certain valves. This modification has no adverse effect on the safety analysis.

f.

.- Nonthly Operations R; port June, 1984 Page Five D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

System Acceptance Committee Packages for June,1984: (Continued)

Package Description / Analysis EEAR FC-79-149 Warehouse Decking Fire Hazard.

This modification did not effect a safety related system; therefore, has no adverse effect on the safety analysis.

EEAR FC-79-153 Fil Flow Transmitter Replacement.

This modification replaced existing FW flow transmitters to improve operability. This modification has no adverse effect.on the safety analysis.

E. RESULTS OF LEAK RATE TESTS The Fort Calhoun Station is currently performing B and C penetration tests. A report will be sent out at the end of the refueling outage.

F. CHANGES IN PLANT OPERATING STAFF None G. TRAINING Training for the month of June for operators was directed toward procedure and administrative changes due to the steam generator tube rupture and the trip two, leave two logic for reactor coolant pumps. Maintenance training was held on flood control procedures, crane operation and systems. Three candidates sat for NRC reactor operator license examinations.

H. CHANGES, TESTS AND EXPERIMENTS' REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 None i

i

~

Nnthly Operations Report -

June, 1984 Page Six II. MAINTENANCE (Significant Safety Related)

A report will be submitted at the end of the refueling outage. .

ste .

W. Gary Gates Manager Fort Calhoun Station s

4 t g

\

i i % i ,

i

% ,. )

i

\

i 4

~

~, \

m h

i t

I k

a e

e f 3

'[

., ,y .~. 1' Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 402/536 4000 July 9, 1984 LIC-84-213 Mr. Richard C. DeYoung, Director Office of Inspection and Enforcement U. S. Nuclea r Regulatory Commission Washington, D.C. 20555

Reference:

Docke t No. 50-285

Dear Mr. DeYoung:

June Monthly Operating Report Please find enclosed ten (10) copies of the June Monthly Operating Report for the Port Calhoun Station Unit No. 1.

Sincerely, AM R. L. Andrews Division Manager Nuclear Production RLA/TPM:jmm Enclosures cc: MRC Regional Office Office of Management & Program Analysis (2)

Mr. R. R. Mills - Combustion Engineering Mr. T. F. Polk - Westinghouse Nuclear Safety Analysis Center INPO Records Center NRC File k

9 45 sit 4 tmployment wan tauai opportunay I Mate remale