ML19296D372
| ML19296D372 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1979 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19296D371 | List: |
| References | |
| NUDOCS 8003040443 | |
| Download: ML19296D372 (12) | |
Text
,
O COOPER NUCLEAR STATION BROWNVILLE, NEBRASKA ANNUAL OPERATING REPORT JANUARY 1, 1979 THROUCH DECEMBER 31, 1979 USNRC DOCKET 50-298 L-8008040
/
TABLE OF CONTENTS SECTION PAGE NUMBER I. PERFORMANCE CHARACTERISTICS AND TESTS 1
Fuel Performance 2
Vessel Transient Condition Events 3
II. FACILITY CHANGES REPORTABLE UNDER 10CFR50.59 5
III. PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION 9
c i
I.
PERFORMANCE CHARACTERISTICS AND TESTS 1
FUEL PERFORMANCE Of f-gas activity in the January 1 through April 7,1979 operational period showed no increases indicative of additional fuel failures beyond those reported in the Cooper Nuclear Station (CNS) Annual Operating Report for 1978. The off-gas activity level continued to decrease from January 1 to April 7, 1979 with the release rates being well within the limits specified in the CNS Technical Specifications.
During the period f rom April 7 ' ' rough dby 6, 1979, the reactor was shut down and the reactor vessel disassembled for the scheduled refueling and maintenance outage.
The core was rearranged as per the fuel loading plan developed by General Electric for Cycle V; 164 spent fuel assem-blies were removed and replaced with 112 new fuel assemblies and 52 used fuel assemblies.
These 52 used fuel assemblies were initial cycle fuel assemblies removed from the reactor during the Cycle II refueling.
In concurrence with General Electric, sipping for leaking fuel assemblies was not warranted due to the low off gas activity. After the reactor core loading was completed, the fuel loading was verified as correct in accordance with the General Electric loading plan for Cycle V and the results recorded on video tape.
On May 6,1979, following completion of the NRC review and approval of the Cycle V licensing submittal, the reactor was started up and the startup physics test program was initiated.
One hundred percent thermal power was initially achieved for Cycle V on June 12, 1979.
From May 6 through December 31, 1979, a gradual increase in of f-gas activity tras monit o red.
This activity, however, was approximately equivalent to the off-gas activity at the beginning of 1979 and indicates no significant chs ge in the number or severity of leaking fuel assemblies in the reactor.
Comparisons of the actual control rod density during the period January 1 to December 31, 1979, to the control rod density predicted by computer prograns at various core average exposures indicated reactivity anom-alles less than 1% AK/K.
The startup physics test program was completed on June 19, 1979.
Notification of test completion was suomitted to the NRC on June 21, 1979.
2
VESSZL TRANSIENT CONDITION EVENTS No operational transients more severe than the transients evaluated in fatigue usage calculations as described in CNS Calculation Book 8.40-21 are identified during this report period.
Reactor Coolant Pressure Boundary thermal and pressure cycles are sum-marized as follows:
NOTE: All transients started from approximately rated pressure and temperature.
MINIMUM TE"PERATURE DATE OF THE MODERATOR (OF) 1 5-6-79 120 2
5-9-79 337 3 5-24-79 208 4
5-26-79 180 5 8-10-79 160 6 9-14-79 118 7 11-20-79 100 There were seven (7) thermal-pressure cycles on the Reactor Coolant Pressure Boundary, six (6) were full cycle (from rated pressure and temperature to vessel vented), during the other cycle the vessel temp-erature did not go bel.ow 337 F as indicated above.
3
VESSEL TRANSIENT CONDITION EVENTS (DESIGN FATIGUE USAGE)
REPORT PERIOD TOTAL EVENTS DESIGN EVENTS TO DATE EVENTS Normal Startup (100 F/hr) 7 93 120 50% Power Operation 14 57 14,600 (reduction to)
Rod Worth Tests 0
3 400 Loss FW Heaters 257. Turbine Trip 0
6 10 FW Heater Bypass 0
1 70 Loss FW Pumps (Scraes) 1 5
10 Turbine Generator Trip 1
17 40 Reactor Overpressure 0
0 1
Safety Valve Blowdown 0
0 2
All Other Scrams 5
43 147 Improper Start Cold RR Loop 0
0 5
Sudden Start Cold RR Loop 0
0 5
Normal Shutdown 4
32 118 Hydrostatic Pressure Tests 0
1 3
(1563 psig)
Hydrostatic Pressure Tests 1 0 1025 8 @ 1025 130 (1250 psig) 4
O II.
FACILITY CIM GES REPORTABLE CIDER 10CFR50.59 5
REPORTABLE MINOR DESIGN CHANGES (MDC) COMPLETED IN 1979 FDC 77-37 Component:
RHR System - Loops A and B
==
Description:==
This SDC authorized installation of isolation valves in the steam lines to the RHR heat exchangers.
These valves permit isolation of the heat exchangers for maintenance.
Operation of the RHR System is not affected by this change.
MDC 78-42 Component:
Motor Control Center K (MCC-K)
==
Description:==
This MDC provided for the addition of the plant security system load to the MCC-K electrical distribution panel.
The chaage puts the security system electrical load on a protected power supply.
The MCC-K panel has sufficient capacity to accommodate the additional load and the security system load is similar to other loads on the panel.
SDC 78-62 Component:
Feedwater Control System Steam and Level Instruments
==
Description:==
This MDC authorized the removal of density compensation from the steam flow and reactor level instruments used to control the feedwater flow (and hence reactor icvel).
This change could not cause reactor level variances larger than the acceptable range of the control system.
The new design reduces the complexity of the system, thereby reducing system outages for maintenance and repair.
MDC 77-69 Component:
Refueling Platform
==
Description:==
This MDC provided for the installation of a load cell with readout on the monorail hoist on the refueling platform. The load cell provides a method for measuring the actual force exerted by the hoist when removing neutron detectors from the core.
This component is a backup to the installed limit switches which prevent overload of the hoist.
6
MDC 78-25 Component:
4160V AC Buses IF and 1G
==
Description:==
This MDC authorized installation of time-delayed under-voltage relays to the 4160V AC buses IF and 1G.
These relays protect the bus loads from prolonged low voltage conditions (less than 3600V AC for greater than 10 seconds) which could affect the operation of electric motors supplied by the buses.
These relays are a second level of protection to the undervoltage relays which trip the loads immediately when bus voltage reaches 2900V AC.
MDC 79-37 Component:
Centrol Room Ventilation Radiation Monitor
==
Description:==
This MDC provided for the relocation of the radiation sample probe in the control room ventilation system.
Although the probe is required to sample moving air, the previous location was not in a moving air stream.
Con-sequently, the effectiveness of the instrument was re-duced.
The new location for the probe is in a moving air stream.
All requirements of the appropriate standards are met.
7
REPORTABLE CIM GES TO FIRE PROTECTION COMMITTMENTS In April 1977, Nebraska Public Power District (NPPD) committed to familiarization tours and radiation training for local fire department personnel.
This was done to justify a 3 man on-site fire brigade.
NPPD has subsequently committed to a 5 man on-site fire brigade to satisfy current NRC requirements.
Consequently, training of off-site personnel has been reduced to that necessary to fight fires outside the plant.
Since the NRC has agreed that the 5 man fire brigade is adequate and since we have committed to a 5 man fire brigade, the change does not involve an unreviewed safety questien.
The change does not alter the plant as evaluated in the FSAR nor does it introduce new safety questions.
No change in the Technical Specifications is required.
The above su= mary of a 10CFR50.59 analysis is being reported to document an agreement reached between the Nuclear Regulatory Commission and Nebraska Public Power District during a January 24, 1979 meeting at the NRC Region IV office.
8
III.
PERS0KfEL AND MAN-REM BY WORK AND JOB FUNCTION 9
~..
PERSONNEL A';D MA'!-REM BY WORK AND JOB FUNCTION 1979 Number of Personnel Total Man-Rem
( > 100 mrem)
Employees (Employees Station Utility Contractor Station Utility Contractor Work and Job Function Employees Employees
& Others
& Other REACTOR OPERATIONS & SURV.
l Maintenance Personnel 2
1.096 Operating Personnel 39 24.234 Health Physics Personnel 13 1
6.111
.393 Supervisory Personnel 9
1 6.411
.116 Engineering Personnel 12 2
7.446
.426 ROUTINE MAINTENANCE Maintenance Personnel 44 70 49.427 55.999 Operating Personnel 3
.961 Health Physics Personnel 8
2.340
. Supervisory Personnel 2
.424 Engineering Personnel 5
1.441 SPECIAL MAINTENANCE Maintenance Personnel 8
27 5.479 21.503 Operating Personnel 2
.330 Health Physics Personnel 8
.745 Supervisory Personnel Engineering Personnel 2
15 1
.550 6.321
.160 WASTE PROCESSING Maintenance Personnel Operating Personnel 8
2.375 Health Physics Personnel 9
.675 Supervisory Personnel 1
.058 Engineering Personnel REFUELING Maintenance Personnel Operating Personnel 27 4.923 Health Physics Personnel 8
.132 Supervisory Personnel 2
.178 Engineering Personnel 3
.562 INSERVICE INSPECTION Maintenance Personnel 8
2.263 Operating Personnel Health Physics Personnel 8
.711 Supervisory Personnel 2
1.223 Engineering Personnel 1
.154 TOTALS Maintenance Personnel 44 97 54.906 80.861 Operating Personnel 39 32.823 Health Physics Personnel 13 1
10.714
.393 Supervisory Personnel 9
3 7.071 1.339 Engineering Personnel 13 15 3
10.153 6.321
.586 t
GRAND TOTALS 118 15 104 115.667 6.321 83.179 i