NLS8800048, Annual Operating Rept for 1987
| ML20147E948 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1987 |
| From: | Trevors G NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| References | |
| NLS8800048, NUDOCS 8803070192 | |
| Download: ML20147E948 (13) | |
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COOPER ~ NUCLEAR STATION-BROWNVILLEl NEBRASKA e
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ANNUAL OPERATING REPORT i
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JANUARY.1, 1987 THROUGH DECEMBER 31, 1987 I
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- i TABLE OF. CONTENTS' SECTION
'PAGE. NUMBER
-I.
PERFORMANCE CHARACTERISTICS AND TESTS 1
Fuel Performance 2:
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MSV and MSRV Failures and Challenges--
3 Reportable-Special Frocedures/Special Test Procedures 4
II.
FACILITY CHANGES REPORTABLE UNDER 10CFR50.59 5
III.
PERSONNEL'AND MAN-REM BY WORK AND JOB FUNCTION 9
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PERFORMANCE CHARACTERISTICS AND TESTS i
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$UEL PERFORMANCE 1
During the period from January 1, 1987, through January 3, 1987, the reactor remained shut down, nas the scheduled refueling and maintenance outage continued.
On January 3, 1987, following NRC review and approval of the-Cycle XII Licensing Submittal, the reactor was started up.
One hundred percent thermal-power was initially achieved for Cycle XII on January 22, 1987.
The startup physics test program was completed on January 26, 1987, with notification of test' completion submitted to the NRC on February 5, 1987.
Off-gas activity in the January 3 through Dacarber _31, 1987, operational period continued at essentially secady-state levels.
Reactor coolant dose equivalent I-131 equilibrium values and off-gas release rates were maintained well within the limits specified by the CNS Technical Specifications.
Comparisons - of the actual control rod densities during the period January 3 to December 31, 1987, to the control rod densities predicted by computer program calculations at various core average exposures, indicated no reactivity anomalies of 1% Ak/k or greater.
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MSV AND MSRV FAILURES AND CHALLENGES s,
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-(Ref: 'NUREG-0737,LAction Item'II.K.3.3)
. There were 'no. ch'a11enges or failures of any ' safety or' relief valves.- during.
1987.
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! REPORTABLE SPFCIAL' PROCEDURES /SPECIAL TEST PROCEDURES There were no' reportable Special Procedures /Special Test Procedures. conducted
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n II.
FACILITY CHANGES REPORTABLE UNDER 10CFR50.59 f
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DESIGN CHANGES COMPLETED IN 1987
.DC 85-096 COMPONENT:
METEOROLOGICAL COMPUTER PMIS-UPS SYSTEM POWER FEED DESCRIPTION:
The purpose of this DC was to relocate the power feed to. the Meteorological Computer from CCPlB,.to the PMIS-UPS System which incorporates a battery back up design feature.
This change put the Meteorological Computer on the same source as the PMIS equipment to preclude the loss of the Meteorological Computer System due to low or loss of voltage to CCPlB.
SAFETY ANALYSIS:
The Meteorological Monitoring System is not safety related and is not required for safe shutdown.
Modifications to safety related equipment included conduit and cable tray supports only.
With regards to support design, seismic Class 1 requirements were met.
The circuit design, including conduit was in routing. and separation of lE and non-lE circuits, accordance with the requirements prescribed in IEEE Standard 384-1981.
Cabling met the requirements of IEEE Standard 383.
Existing fire wall conduit penetrations were used; therefore, installation of additional fire semis was not required.
DC 86-096 COMPONENT:
MCC QUALIFICATION UPGRADE DESCRIPTION:
This Design Change (DC) provided for the replacement of the internal components of selected Motor Control Centers (MCC's),
located in the Reactor Building, with environmentally and seismically qualified components.
The work performed under thic DC consisted of:
Replacement of internal components including the fused disconnect switches, starters, overload relays, control power transformers, fuses, and wiring including the wiring between the compartments and the master terminal blocks in the top hat of the MCC, of all compartments of McC's R, RB, and Q.
Replacement of five circuit breakers in MCC's K and S with fused disconnect switches.
Replacement of the fuses in the fused disconnect feeder fo MCC Q.
It was the intent of this upgrade to standardize the fuses in these MCC's to Bussman, Low Peak, Type LPS fuses. However the feeders to MCC RA have Bussman Type FRS fumes to coordinate with the existing fuses in MCC RA. _ _ _ - _ _ _ _
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' SAFETY-ANALYSIS:
This DC required modifications to safety related equipment and-was considered essential.
The_ implementation of this DC does'not in any way degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety.
It was the intent of this DC to upgrade plant safety by installing qualified ' equipment to further ensure the operability of safety related equipment following a DBE.
The installation of the fused disconnect switches in place of the circuit breakers for MCC feeders increased the reliability of the electrical ' system.
All equipment installed for this DC was designed to meet required standards and environmental and seismic criteria.
This DC does not increase the probability or consequences of an 1
accident, nor does it decrease the safety margin as defined in the basis for a Technical Specification.
DC 86-082 COMPONENT:
RCIC CABLE ISOLATION DESCRIPTION:
The purpose of this Design Change was to isolate Cables RC40 and RC72, associated with reactor vessel low water level, and RC41, associated with reactor high water level.
These changes were determined to be necessary following various Appendix R evaluations and resultant (potential) fire induced spurious reactor high water level operations.
Cable RC41 provides a
indication signal back to Panel 9-30.
Cables RC40.and RC72 brought a reactor low water level indication signal back to Panel 9-30.
The isolation was accomplished by installing one six stage isolation switch on Relay Panel 9-30 and wiring the normally closed contacts of the isolation switch in the circuits to prevent the spurious energization of Relays 13A-K2, 13A-K3, 13A K17, and 13A-K38X, also located on Panel 9-30.
During postulated fire scenarios involving Cables RC40, RC41, and RC72, the isolation switch will be operated from NORMAL to ISOLATE position; opening-the isolation contacts and preventing spurious energization of the above relays.
Spurious energization of the relays could have resulted in undesirable RCIC System valve operation which may have limited adequate l
cooling water being provided to the reactor. Operation of the new isolation switch is annunciated in the control Room (Panel 9 4).
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' SAFETY ANALYSIS:
This' Des"ign Change involved the addition of an isolation. switch to the'RCIC System.
The addition of-the new isolation switch
-did not increase the probability of occurrence or the consequences of an -accident or malfunction ~ of : equipment
-important-to safety-as previously evaluated in the USAR because the. functional. configuration of the RCIC - System remained unchanged.
The undesirable spurious energization of relays 13A-K2, 13A-K3, 13A-K17, and.13A-K38X.were mitigated, increasing the probability for safe shutdown during postulated fire scenarios.
DC 86-101 i
COMPONENT:
MAIN STEAM LINE DRAIN PIPING REPIACEMENT DESCRIPTION:
This Design-Change authorized the replacement of a portion of the main steam line drain piping.
The existing carbon steel piping from the downstream side of Valve MS MOV-79 to. the condenser, approximately 300 feet was. replaced with stainless steel piping.
The one inch piping downstream of MS-RO-130 was also replaced with stainless steel piping. Due to erosion, the existing carbon steel piping had become thin walled,- and in at least one location, the piping had failed, resulting in leakage.
Use of stainless steel piping will resist erosion, prevent steam leaks, and improve system reliability.
In addition, the drain lines were connected to the extraction.
steam lines leading to Heaters SA and SB.
This utilizes the heat energy of the steam and water mixture in the drain lines for feedwater. heating. Two valves were used to direct drainage to either the SA and SB heaters or to'the condenser.
During normal power operations drainage will be through the normally
-open valve to the heaters and drainage to the condenser will be blocked by the normally closed valve.
SAFETY l
ANALYSIS:
No systems important to nuclear safety were affected by this i
activity.
Stainless steel piping, which is resistant to erosion was used to replace the existing carbon steel piping, which had suffered extensive erosion damage.
The possibility I
of steam leaks from this line is reduced.
The reliability of l
the specific line, main steam drain, will be increased.
Part of the replacement activity necessitated removal and reinstallation of pipe through a secondary containment penetration.
To eliminate any safety concerns, those activities which breached secondary containment were limited to when secondary containment was not required.
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III.
PERSONNEL AND MAN REM BY WORK AND JOB FUNCTION L
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PERSONNEL AND' MAN-REM BY WORK AND JOB' FUNCTION-Number of Personnel Total Man-Rem
(> 100 mrem)
.-Station Utility Contractor.
Station
. Utility Contractor Werk snd Job Function
' Employees Employees & Others Employees Employees & Other REACTOR OPERATIONS & SURV.
M:intenance Personnel 12 0
1 0.156 0.000 0.010 Oparating Personnel-34 0
0 11.711 0.000 0.000 H1alth Physics Personnel-21 0
0 6.163 0.000 0.000 Supervisory Personnel 10 0
0 2.367 0.000_
0.000 Engineering. Pers onnel 3
2 5
1.184 0.005 0.541 ROUTINE MAINTENANCE Mnintenance Personnel 67 0
18 36.592 0.000 8.021 Opsrating Personnel 3
0 0
0.033 0.000 0.000 Haalth Physics Personnel 15 0
0 8.871 0.000 0.000 Supervisory Personnel 4
0 1
0.655 0.000 0.191 Engineering Personnel.
0 10 3
0.000 2,437 0.220 SPECIAL HAINTENANCE Maintenance Personnel 1
0 1
0.005-0.000 0.137 Oparating Personnel 1
0 0
0.015 0.000 0.000
'Hosith Physics Personnel 7
0 0
0.459 0.000 0.000 Supervisory Personnel 1
0 0
0.122 0.000 0.000 Engineering Personnel 1
0 0
0.089 0.000 0.000
' WASTE PROCESSING M intenance Personnel 0
0 0
0.000 0.000 0.000-Opsrating Personnel 11 0
0 4.446 0.000 0.000 Hsolth Physics Personnel 12 0
0 3.235 0.000 0.000
-Supervisory Personnel-2 0
0 0.008 0.000 0.000
-Engineering Personnel 0
0 0
0.000 0.000 0.000 REFUELING Mnintenance Personnel 3
0 0
0.028 0.000 0.000 Op; rating Personnel 1
0 0
0.005 0.000 0.000 H221th Physics Personnel 0
0 0
0.000 0.000 0.000 Supervisory Personnel 0
0 0
0.000 0.000 0.000 Engineering Personnel 0
0 0
0.000 0.000 0.000 INSERVICE INSPECTION Maintenance Personnel 0
0 1
0.000 0.000 0.108 Opsrating Personnel 1
0 0
0.009
-0.000 0.000 Haelth Physics Personnel 0
0 0
0.000 0.000 0.000 Supervisory Personnel 1
0 0
0.001 0.000 0.000 Engineering Personnel 0
0 0
0.000 0.000 0.000 TOTALS Maintenance Personnel 67 0
29 36.781 0.000 8.296 Opsrating Personnel 38 0
0 16.219 0.000 0.000 Hasith Physics Personnel 21 0
0 18.728 0.000 0.000 Supervisory Personnel 10 0
1 3.153 0.000 0.191 Engineering Personnel 3
10 5
1.273 2.442 0.761 GRAND TOTALS 139 10 25 76.154 2,442 9.248 w
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GENERAL OFFICE Nebraska Public Power _ District
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NLS8800048 February 29, 1988 Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Annual Operating Report Cooper Nuclear Station NRC Docket No. 50-298, DPR 46 In accordance with Paragraph 6.5.1 of the Cooper Nuclear Station Technical Specifications, the Nebraska Public Power District submits the Cooper Nuclear Station Annual Operating Report for the period of January 1, 1987, through December 31, 1987.
We are enclosing one signed original for your use and, in accordance with 10CFR50.4, are transmitting one copy to the NRC Regional Office, and one copy to the NRC Resident Inspector for Cooper Nuclear Station.
Should you have any questions or comments regarding this report, please contact me.
Sincerely, fif7 eorg A. Trevors Division Manager of Nuclear Support GAT /gmc:mh21/6(PPGC2H)
Enclosure cc: NRC Regional Office Region IV NRC Resident Inspector Cooper Nuclear Station Division Manager of Nuclear Operations w/l Enclosure Cooper Nuclear Station f&
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