NLS8900094, Annual Operating Rept for 1988
| ML20235T349 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1988 |
| From: | Trevors G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS8900094, NUDOCS 8903080248 | |
| Download: ML20235T349 (38) | |
Text
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2 COOPER NUCLEAR STATION BROWNVILLE, NEBRASKA ANNUAL OPERATING REPORT JANUARY 1, 1988 THROUGH DECEMBER 31, 1988 USNRC DOCKET 50-298
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TABLE OF CONTENTS SECTION PAGE
-1.
PERFORMANCE CHARACTERISTICS.
1 Fuel Performance 2
MSV and MSRV Failures and Challenges 3
II.
FACILITY CHANGES,. TESTS, OR EXPERIMENTS REPORTABLE 4
UNDER 10CFR50.59 Reportable Special Procedures /Special Test Procedures.
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Reportable Design Changes 17 III.
PERSONNEL AND MAN-REM EXPOSURE BY WORK AND JOB 34
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PERFORMANCE CHARACTERISTICS I
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FUEL PERFORMANCE Cycle 11 operation was interrupted on January 28, 1988, as a result of APRM High flux Scram. (Reference LER 88-002).
The unit was restarted and continued operation through March 5, 1988.
Off-gas activity continued at essentially steady-state levels with reactor coolant dose equivalent I-131 equilibrium values and off-gas release rates maintained well within the limits specified by the CNS Technical Specifications.
Comparisons of the actual control rod densities to the control rod densities predicted by coinputer program calculations at various core average exposures indicated no reactivity anomalies of 1% Ak/k or greater.
During the period from March 5 through June 17, 1988, the reactor was shutdown and the reactor vessel disassembled for the scheduled refueling and maintenance outage. An in-vessel fuel shuffle was performed which included replacement of 124 fuel assemblies. With the concurrence of General Electric, it was decided that sipping for leaking fuel assemblies was not warranted due to the low off-gas activity.
Cycle 12 operation commenced with initial reactor startup on June 17, 1988, and 100 percent thermal power was initially achieved on July 4,1988. Operation was interrupted by a forced outage on July 15, 1988, due to a. hot spot and burned gasket on Isolated Phase Bus Duct (Reference LER 88-019). The unit was restarted on July 16, 1988.
Operation was again interrupted as a result of a Main Steam Line High Radiation Monitor spike and subsequent scram on August 15, 1988 (Reference LER 88-021). The unit was restarted on August 17, 1988, and continued operation through December 31, 1988.
The startup physics test program was completed on August 3, 1988, with notification of test completion submitted to the NRC on September 15, 1988.
Off-gas activity continued at essentially
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steady-state levels with reactor coolant dose equivalent I-131 equilibrium values and off-gas release rates maintained well within the limits specified by the CNS Technical Specifications.
Comparisons of the actual control rod densities 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 FAILURE CHALLENGES (Ref: -NUREG-0737, Action Item II.K.3.3)
There were' three challenges to the. relief valves during the August 25, 1988, scram. All three valve actuations were satisfactory.
There were no failures.
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II, FACILITY CllANGES, TESTS, OR EXPERIMENTS, REPORTABLE UNDER 10CFR50.59
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REPORTABLE SPECIAL PROCEDURES /SPECIAL TEST PROCEDURES SP 85-014-TITLE:
CNS/ Nuclear Packaging On-Site Radwaste Processing DESCRIPTION: This Special Procedure (SP) provided for the temporary on-site processing of Radwaste spent resins and sludge with the Nuclear Packaging, Inc. (NuPac) Dewatering System.
SAFETY ANALYSIS':
Special Procedure 85-014 provided a method of disposing of wet solid wastes which allowed for a greater margin of safety compared to the existing radwaste solidification equipment, by reducing personnel exposure, waste volume, and providing greater assurance of meeting 10CFR61 requirements.
Therefore, the overall plant safety was increased.
SP 87-002 TITLE:
Testing of HPCI Overspeed Tappet Assembly DESCRIPTION: This Special Procedure was conducted to properly document the testing of the High Pressure Coolant Injection (HPCI) overspeed unit.
The testing was conducted every two weeks for a 3 month period.
SAFETY ANALYSIS:
Performance of this Special Procedure was done in conjunction with the weekly operation of the HPCI auxiliary lube oil Pump S.P.6.4.3.2.
This SP did not require abnormal operation of any plant system or procedures, and did not alter any plant equipment.
Safety was enhanced through the verification of the operability of the HPCI overspeed trip device.
E,P 88-001 TITLE:
Fire Pump D and E Testing DESCRIPTION:
The purpose of this Special Procedure was to determine if the existing 8" Valve FP-V-657 can be used to equalize pressure across fire pump discharge Valves FP-V-561 and FP-V-564 during the weekly fire pump operability tests.
SAFETY ANALYSIS:
The Fire Protection System is not essential although it performs an important function.
This Special Procedure involved minor changes to S.P. 6.4.5.3 with local valve manipulations. Altering the existing Surveillance Procedure to obtain additional engineering data did in no way create a nuclear safety hazard.
Station Operators were present at all times during the test to return the Fire Protection System to a normal alignment if problems had occurred. l~
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s SP 88-217 TITLE:
Reactor Recirculation (RR) Operation Above Minimum Flow With Less Than 20% Feedwater Flow.
I DESCRIPTION: This procedure provided instructions, in addition to those provided in System Operating Procedure 2.2.68 Reactor Recirculation System, and General Operating Procedure 2.1.4 Normal Shutdown From Power, for the operation of the Reactor Recirculation (RR) System with the interlock bypassed which restricts RR to minimum speed when feedwatet flow is below 20% of rated flow.
This procedure will facilitate troubleshooting of the Reactor Recirculation Feedwater Control (RRFC) system.
SAFETY ANALYSIS:
Bypassing the 20% feedwater flow interlock is enveloped by failure analyses in the USAR which determines that any failure of the Recirculation Control System cannot result in exceeding a safety design basis. Additional assurance that system safety functions of the RR and Vessel Internal Systems would not be adversely affected was provided by imposition of limitations on RR System operation based upon an evaluation by the NSSS vendor (G.E.). This activity did not represent an unreviewed safety question or a violation of the Technical Specifications. This SP was performed with the reactor in a cold shutdown condition.
STP 86-014 TITLE:
Testing Diesel Generator Lube Oil Filter Gasket DESCRIPTION: The purpose of this Special Test Procedure was to install and evaluate tha performance of a newly designed diesel generator lube oil filter gasket.
With this test procedure, information was gathered to determine if the newly designed gasket would reduce the alignn'ent problems that occur during installation, which will result in a better seal.
SAFETY ANALYSIS:
The diesel generator lube oil filter and its components are classified as safety related. The installation of the redesigned gasket in the diesel generator lube oil filter did not create a nuclear safety hazard other than what has been previously evaluated.
The safety objective of the diesel generator was not decreased with the installation of the redesigned gasket.
The operation of the Diesel Generator System was not changed.
STP 86-021 TITLE:
Feedwater Heater Drain Evaluation DESCRIPTION:
The objective of this Special Test Procedure was to evaluate the impact on plant chemistry if forward pumped heater drains are ___
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d installed at Cooper Nuclear Station.
Additionally, a complete corrosion product balance was made around the heater drain system for determining source terms and sinks for these ' corrosion products.
SAFETY ANALYSIS:
This procedure involved portions of the Condensate Drain System which is non-essential.
This Special Test Procedure did not.in any way degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety.
All work performed under this test was done in accordance with procedures and-practices normally performed in the routine operation of the plant.
STP 87-001 TITLE:
Pressurization Test of Control Room DESCRIPTION:
The purpose of this STP was to determine the positive static pressure in the Control Room / Cable Spreading Room envelope relative to the surrounding areas and to ambient, and to determine the amount of air needed to pressurize the Control Room / Cable Spreading Room envelope to specific static pressure levels. This information will be used in writing the Basis of Design Document (BODD) for the second phase of the Control Room HVAC modification. The test took place in two modes of operation of the system. First, in the normal mode to determine the amount of air required to pressurize the room to various levels. The second, in the emergency mode to determine the effectiveness of the existing system.
SAFETY ANALYSIS:
This special test procedure did not in any way degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. The test was conducted with the system first in-the' normal mode, then in its standard emergency mode.
The system remained capable of being switched to the emergency mode of operation and be fully operable if needed as directed.by the Shift Supervisor. This STP did not require abnormal operation of any plant system or procedure and did not introduce any plant equipment alteration.
Therefore, the affect on overall plant safety was not changed.
STP 87-004 TITLE:
Feasibility of Conducting HPCI Surveillance Testing at Referenced Pump Speed DESCRIPTION: The purpose of this Special Test Procedure was to determine the feasibility of conducting quarterly High Pressure Coolant Injection l
(HPCI) pump surveillance testing at referenced pump speed and l
discharge head for Inservice Testing versus the existing method I
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of testing at reference pump flow and total discharge head.,
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SAFETY ANALYSIS:
LThe1 High Pressure Coolant Injection System is, classified as a.
safety-related system.
Altering.the. existing---surveillance procedure to obtain additional engineering data did in no way
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create a nuclear safety hazard because the safety objective of the HPCI system remained intact and no changes in the operation of the HPCI system were made.
STP 87-005 and'RevisioD_1
- TITLE':
Bleed'and Lockup Rates on Snubbers DESCRIPTION: The objective of this STP was to test three' hydraulic snubbers for lockup and bleed rate at various-temperatures.
The periodic.
functional. testing as required by S.P. 6.3.10.9.1
" Snubber Operability Test" is usually performed at atmospheric temperature -
and the lockup and bleed rates are analytically compensated for temperature.
This is not indicative as to how the snubbers function at actual operating conditions. Therefore, this STP was
.to provide actual test data in which the' lockup and bleed rates found during functional testing can be corrected to actual operating conditions.
' SAFETY ANALYSIS:
In the past, several snubbers tested at atmospheric, temperatures were found to have satisfactory values for lockup and bleed rate.
. However, due to the analytical temperature compensation for. actual-operating conditions. these snubbers failed.
Therefore, three 2b" x 5" stroke Crinnell hydraulic snubbers were sent to a testing laboratory and tested for lockup and' bleed rates at controlled temperatures ranging form 60*F to 160*F at 20*F increments.
The three snubbers were rebuilt snubbers and set as close as possible to 8 ipm for the lockup velocity and 4 imp for. the bleed velocity.
After the test data was received, CNS Plant Engineering generated a relation for temperature to lockup ~and bleed rate.
These correction factors were used in S.P.. 6.3.10.9.1 for evaluating actual operating conditions of the snubbers installed in the plant.
This test did not constitute an unreviewed safety question as no change or modification.was performed. The scope of this work was testing only and was performed with the snubbers removed from the plant.
'STP 87-007 TITLE:
Conducting RCIC Surveillance Testing at Referenced Pump Speeds DESCRIPTION: The purpose of this Special Test Procedure (STP) was to determine the feasibility of conducting quarterly Reactor Core Isolation Cooling (RCIC) pump surveillance testing at a referenced pump speed and at a required pump discharge head for Inservice Testing. This
.STP also determined if the RCIC turbine speed will remain constant as the system head is changed when RCIC flow indicating controller RCIC-FIC-91 is in manual.. _ _ - _ _. - _ _ - _ _ - _ _ _ _ - -
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SAFETY.
ANALYSIS:
The RCIC system is classified as a safety-related system. Altering-the existing surveillance procedure to obtain additional engineering data did in no way create a nuclear safety hazard because the safety objective of the RCIC system could still be met and no changes in the operation of the RCIC system were made.
.STP 87-009 TITLE:
Air Flow Increas.e to Critical Switchgear Rooms DESCRIPTION: The purpose of this STP was to examine the effect of a larger cooling capacity in the Critical Switchgear Rooms (CSGR's) ventilation system.
There was a concern.with high temperatures near design limits in the Critical Switchgear Rooms. By increasing -
air flow through the room, a larger cooling capacity was gained
. without : a major effect on the rest of the - system.
This STP collected data to write a design change which will increase the ventilation air flow through CSGR's.
SAFETY ANALYSIS:
STP 87-009 tested duct air flow in the Critical Switchgear Rooms and no permanent modifications were made to. the system.
The original features of the duct system were restored after the test.
During the test fire protection capability for detection and suppression remained intact.
Therefore, the affect on overall plant safety was not changed.
STP 87-011 TITLE:
Post-LOCA Service Water System Flow Test l
DESCRIPTION: The purpose of this Special Test Procedure was to gather data for Service Water System operation in a lineup which simulates loss of power conditions when only one service water pump is available.
This data was used to determine heat exchanger, pump, and flow path performance in response to the recent Safety System Functional Inspection (SSFI).
Each service water pump was tested in a condition which simulates required service immediately following a LOCA and ten minutes after a LOCA without offsite power available.
Pump flow in these lineups was varied in order to develop pump \\ system head loss curves.
Data for REC and RHR heat exchangers was gathered while these heat exchangers were operated in their normal modes.
SAFETY ANALYSIS:
Due to the lineups required for the test, SW pump flow tests were performed while the plant was shut down. However, all components served remained operable to perform their intended function.
During parts of the STP, one diesel generator was operated in accordance with its surveillance procedure.
Other equipment affected was operated in modes covered by CNS operating procedures.
The purpose of this STP was to gather data for Service Water System operation in a lineup which simulates a LOCA and loss of offsite _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
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power conditions. Actual abnormal test lineups were performed' kith the plant shut down and cooled down.
Systems remained operable to perform their intended functions.
No permanent changes to systems resulted from this STP. Therefore, the affect on overall' plant safety was not changed.
STP 87-014 TITLE:
Measurement of Plant 120 VAC Loads DESCRIPTION: 'lhe purpose of this test was to determine the actual loads and voltages for the transformers that feed critical Distribution Panels CDPlA and CDP 1B during normal operating conditions with the plant output greater than 20MWe. This STP provided information to produce a voltage profile study of CNS to ensure that.the plant electrical distribution system will he capable to mitigate a postulated accident scenario. In addition this test provided the actual loads and voltages on critical control Panels CCPlA, CCPIB, CPP, and CPP-2 with the plant in cold shutdown.
To determine loading, measurements of panel bus voltage and input current to the panel were taken, also selected output feeder cable currents were taken.
SAFETY ANALYSIS:
This STP did not require abnormal operation of any of the plant systems and did not introduce any plant equipment alterations.
This STP provided the steps for basic electrical voltage and current measurements at 120 VAC critical instrumentation and control power panels, and the transformers that feed these panels, one division power source at a time.
The existing electrical terminations were not disturbed during the measurements.
STP 87-016 TITLE:
Torus Water Level Compensation Curve Verification DESCRIPTION: The STP was written to determine the affect the core spray pumps have on the torus ' level indication during core spray operation.
The point readings taken were used to determine whether the graph developed in Calculation 87-031 needed to be revised to reflect a more accurate conversion graph. This graph is used by operators to compensate the torus level indication while the core spray pumps are being used.
SAFETY ANALYSIS:
This STP inserted two steps into Procedure 6.3.4.1 "CS Test Mode Surveillance Operation" which involves the core spray systems, thus, this STP involved essential safety related equipment. The steps inserted instruct operations personnel to control the amount of flow going through the CS pump to different flow rates. Since only extra readings were taken from with the Control Room and on Plant Management Information System (PMIS) during the performance of S.P. 6.3.4.1, no adverse effects were placed on plant equipment - _ - _
3 and personnel. Nuclear safety considerations did not differ from those taken during performance of S.P. 6.3.4.1.
STP 87-017 TITLE:
Vessel Beltline Outside Wall Temperature DESCRIPTION: The purpose of this Special Test Procedure was to confirm that the thermocouple (TC 2-3-69 J1) located 36-inches below the feedwater nozzle, accurately reflects outside wall temperature at the vessel beltline region during shutdown conditions. Confirmation that the existing thermocouple correlates to actual vessel wall temperature would allow designation of the thermocouple to monitor vessel wall temperature during an in-service hydrostatic or leak test.
This was accomplished by using a contact thermocouple and a hand held indicator at two locations on the outside vessel wall at the beltline.
This data was compared to the existing thermocouple data to obtain the correlation.
SAFETY ANALYSIS:
STP 87-017 measured the temperature of the vessel outside wall at two locations during shutdown. A surface contact thermocouple and hand held indicator were used. This required the removal of two panels of vessel insulation and in no manner required a deviation from procedures, alteration of equipment or special system line-ups.
Therefore, this activity did not increase the probability of occurrence or the consequences of an accident, create a possibility for an accident or malfunction of a different type than analyzed nor reduced any Technical Specifications margin of safety for Cooper Nuclear Station.
STP 87-020 TITLE:
Moisture Separator Differential Pressure Measurements DESCRIPTION: The purpose of this Special Test Procedure (STP) was to determine the pressure drop across each of the four (4) moisture separators located in the Turbine Building during normal plant operation.
These pressure measurements were used to verify that the pressure drop across each moisture separator had not increased after modification, which is a performance guarantee for Contract 87-llA,
" Moisture Separator Upgrade".
This STP was also part of the operational Acceptance Test for DC 87-123, " Installation of Moisture Separator Differential Pressure Measuring Equipment".
SAFETY ANALYSIS:
This procedure involved portions of the Main Steam System which are non-essential. This Special Test Procedure did not in any way degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. This test was conducted with the system and plant in their normal mode of operation. This STP did not require abnormal operation of any plant systems or procedures, and did not introduce t.',
plant equipment alteration.
Therefore, the affect on overall plant. safety was not changed.
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STP 87-023 TITLE:
~ Measurement of. Plant 120 VAC Loads (NBPP).
DESCRIPTION: The purpose of this Special Test Procedure was to determine the
. actual current End. voltage-loads that the No Break. Power Panel.
(NBPP) and UPP-C-1A panel supply.
To. determine ; loading, measurements of panel. bus' voltage'and input current to the: panel l
and' selected output feeder cable currents were'_taken.
This STP'
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- p was' performed to ensure that the NBPP had sufficient capacity to
. connect'the alternate security system feeder.to the panel.
. SAFETY i
i This STP did not requ' re abnormal operation of any plant systems -
ANALYSIS:
or procedures, and 'did= not introduce.any plant equipment alternation.- JThis STP provided the steps for basic electrical
-voltage and. current _ measurements at the No Break. Power. Panel and at pan,els fed by the No Break Power Panel. The existing electrical-
. terminations were not disturbed during the measurements.
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STP 88-001 TITLE:
Testing of 125 VDC HFA Relays in Safety.5ystems DESCRIPTION: STP 88-001 provided necessary data to verify the ability of the Class 1E.125 VDC HFA relays to perform their safety function General Electric Service Information Letter (SIL).,
SIL 44, Supplement.2 (February 13, 1981) and Supplement 4, Revision 1 (March 14,.1984) were issued to recommend the replacement of the original relay with the new " Century Series" type.
This-recommendation was made due to coil insulation degradation problems with the older style relay that were noted by G.E.
The STP removed all 185 Class 1E General Electric DC HFA relays form service momentarily, one at a time, for testing purposes.
.The test injected a low level 1-2 velt signal into the coil of each relay while an oscilloscope provided a signal trace.
Comparing the
- traces of the tested relays with a known good relay signal trace indicated the condition of each relay.
SAFETY ANALYSIS:
All involved relays were rated at 125 volts DC.
Each relay coil was de-energized prior to lead removal.
All testing was accomplished when the affected system was removed from service for system maintenance.
This limited the possibility of inadvertent challenges to engineered safety functions.
Indirectly affected lt systems were also removed from service, if needed to prevent unnecessary actuations. All testing was conducted from the back panel of the relay, thereby eliminating the possibility of inadvertent relay actuation by the testing personnel.
All necessary precautions were listed specifically for each relay on the respective relay data sheet. _
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~STP 88-013 TITLE:
"C" RHR Motor Testing-DESCRIPTION: The purpose of this activity was to obtain sufficient additional'
. engineering data to complete a failure analysis on the "B" RHR pump i motor ' which occurred..in January,; 1988. (LER 88-003).
.Special Maintenance Procedure SMP88-044 removed the "C" RHR pump motor and replaced it with the repaired. "B" motor.
Once: removed the "C"
. motor was transported to the Multi-Purpose Facility (MPF) for both destructive. : testing : in accordance-with' nondestructive. and STP 88-013. Since the "C" motor winding was destroyed.during this-L'
~ testing process, this procedure was meant to provide documentation 1
of the inspection findings and to provide a record of traceability.'
h of the mechanical motor parts when the-final repair takes' place.
SAFETY
. ANALYSIS:
Based on the insulation fretting observed during the recent re-wind of the "B" motor, it was necessary to perform!a number of tests on the "C"
motor to further define the condition of the motor insulation on the balance of the ECCS motors -at Cooper Nuclear Station.
These tests were both non-destructive and destructive
.in nature and were performed under the direction'of the General.
. Electric / San Jose Engineering representative on site.' This testing-process destroyed only the existing winding' system of a' removed-motor and presented no changes.to the mechanical-parts of the equipment. Further planned repair work constituted a modification to - this equipment but that work was not performed under, this
. procedure.
This test did not constitute an unreviewed safety.
question as no change or modification was being performed and the "C"
motor was replaced prior to any destructive testing. of the motor.
This scope of work was inspection / testing only and was performed with the pump motor removed from the plant.
S.TP 88-015 TITLE:
APRM "C" Inoperative Trip Testing
. DESCRIPTION: The purpose of this Special Test Procedure (STP) was to record data which was used in determining why the Average Power Range Monitor (APRM)
"C" Inoperative Trip setpoint was drifting.
This STP monitored and recorded two critical voltages used by the Quad Trip Circuit to create an APRM Inoperative Trip.
Analysis of the recorded data enabled CNS to establish a logical approach in determining the root cause of the problem.
SAFETY ANALYSIS:
This STP was performed while the Reactor Mode Switch was in the Shutdown or Refuel position.
The APRM system is designed to operate with only two channels per trip system.
This STP took APRM "C" out of service by bypassing it, but APRM "A" and "E" were j
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unaltered and remained operable, thereby, ensuring the two channel trip system. Also a provision was included that required this test to be suspended if at anytime during this STP two (2) operable APRM channels in RPS A were not maintained.
SIP 88-201A TITLE:
Reactor Building HVAu Test Requirements DESCRIPTION: The purpose of this special test procedure was to measure the air and water flow rates, temperatures, pressures, fan motor volts, and fan motor amps for the various Reactor Building supply.and exhaust fans (Fan Nos. 1-SF-R-1A-A, 1-SF-R-1A-B, 1-EF-R-1A, and 1-EF-R-1B), ~ the MG set exhaust fans (Fan Nos. 1-EF-R-lC and 1-EF-R-1D) and the fan coolers for the RHR, core spray, and HPCI pump cubicles (Fan Cooler Nos. 1-FC-R-lE, 1-FC-R-lF, 1-FC-R-1G, 1-FC-R-lH, and 1-FC-R-lJ).
This information will be used in writing the Basis of Design Document (BODD) for the Reactor Building cooling system upgrade.
The test was conducted in two modes of operation of the cooling system. First, all measurements were taken with "A" train fans on the cubicle coolers running.
Second, all measurements were taken with the "B" train fans on the cubicle coolers running.
SAFETY ANALYSIS:
This procedure involved the Reactor Building air supply and exhaust system, and the RR (Reactor Recirculation ) MG-set ventilation system which are non-essential systems in addition to the RHR, Core Spray and HPCI cubicles fan-cooler units, which are essential cooling systems.
This procedure did not affect the operational mode of the systems nor indirectly did it have any effect on any safety function for CNS.
The measurements were taken with the non-essential systems operating normally without any disconnection from its in-line operation.
At the same time the essential fan cooler systems (two redundant trains) were started since they are inactive during normal power operations, and continued to operate during the time all measurements were taken. Since the safety mode of the fan coolers is to be operable and/or operating, it is clear that during the measurements period, the fan coolers remained in a safe mode of operation, that is fully operable.
The Reactor Building and RR MG-set ventilation systems are non-essential with the exception of some valves and ductwork required for secondary containment.
However, this STP did not, in any way, degrade or affect secondary containment leak tightm es.
The purpose of this STP was to collect data and in no way did it create a nuclear safety hazard because the safety objective of the systems remained intact.
STP 88-202 TITLE:
Victoreen High Range Containment Monitoring DESCRIPTION: The purpose of this Special Test Procedure was to collect detector sensitivity (A/R/H) and linearity data, perform calibrations, and _ _ _ _ _ _ _ _
_ _ _ = _ _ _
.f-setpoint calculations on the Victoreen High Range Containment Monitors (RMA-RM-40A and RMA-RM-40B),
and. their associated detectors (RMA-RM-40A and RMA-RE-40B).
SAFETY ANALYSIS:
This STP was performed with the reactor in a cold shutdown condition, when the containment monitors are not required to be _
j operable per Technical Specifications. The equipment was returned to normal operation at the completion of this STP, prior to startup. This activity did not impact related personnel, equipment or nuclear safety conditions.
STP 88-205 TITLE:
Jet Pump Operability Data DESCRIPTION: The purpose of this Special Test Procedure was to collect. data which will be used to verify the curves for performing the Jet Pump Operability Surveillance during startup when recirculation.
temperature is below-rated.
SAFETY ANALYSIS:
STP 88-205 did not authorize any equipment alterations or deviations from plant procedures and Technical Specifications.
This test only defined what data needed to be recorded during plant heat-up and normal startup and did not direct any plant actions or operations. Therefore, this STP did not affect any plant design bases -design criteria or safety analysis.
STP 88-206 TITLE:
Security System Copy Contact Function DESCRIPTION: The purpose of this procedure was to determine whether or not the station security computer system's copy contact design would allow one input to a multi-level board to drive a single output on a contact out board.
This test was necessary to determine if this-function can be used in the security system upgrade project.
SAFETY ANALYSIS:
This test in no way affected plant personnel or nuclear safety at Cooper Nuclear Station.
Only the security computer system was affected on a very limited basis for the duration of the test.
This activity did not affect any safety-related equipment nor degraded security coverage at CNS. Necessary security precautions were taken during the test in accordance with the CNS Safeguards Plan and Procedures.
SMP 88-03 TITLE:
Replacement of 480V "F" Transformer Bushings and Flexible Shunts
' DESCRIPTION: The purpose of this Special Maintenance Procedure (SMP) was to provide authorization for the installation of temporary feeders _ _ - _ _ _ _ - _ - _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _.____ _
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and discennect switches in the 480V distribution system for the
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purpose of providing temporary power to selected equipment during the replacement of bushings and flex shunts on the Station Service I
l Transformer (SST) 1F.
In order to. repair the bushing leaks on SST "F", the entire 480V Bus IF (Division I) was de-energized for approximately three days.
Temporary feeders and disconnect switches were installed permitting energization of several Division I MCCs and selected equipment from Division II sources.
SAFETY ANALYSIS:
The installation of the temporary power feeders and disconnect i
switches during the bushing replacement on SST "1F" was performed while the reactor was in a shutdown, cooled down condition with no fuel movements in progress.
During the power outage created by the repair to SST 1F, separation criteria and electrical separation were allowed to be compromised with the use of this detailed procedure and CNS Procedure 0.9, " Equipment Clearance and Release Orders."
The requirements for these criteria are decreased, and the operational and accident concerns defined in the USAR and Technical Specifications are minimized with the reactor in a shutdown and cooled down condition and no fuel movements in progress.
Transfer of loads from Division I to Division II is allowed during maintenance outages on a recurring basis to support the annual Station Battery Test Discharge Surveillance Procedure, and this activity, properly controlled and documented, expanded the extent of how many loads / buses are to be fed from the opposite electrical division.
Therefore, the activities controlled by this Special Maintenance Procedure did not degrade personnel, equipment safety, nor was the capability to achieve and maintain cold shutdown as described in 10CFR Part 50 compromised by the cross-connection of divisional power supplies.
Af ter completion of this SMP all temporary feeders and disconnects were removed and all system line-ups (Div. I/Div. II) were restored to normal prior to moving any fuel. _
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REPORTABLE DESIGN CHANGES l
l DC 82-089 TITLE:
Turbine Building Crane 193 Ton Rerate Modification 4
DESCRIPTION: This Design Change increased the lifting capacity of the Turbine Building crane from 175 tons to 193 tons..This was done through a rerate study conducted by.the crane manufacturer, and this I
I eventually led to the replacement of several crane components which would have been overloaded at 193 tons.
SAFETY ANALYSIS:
This Design Change will prevent damage to equipment and injury to station personnel from operating the Turbine Building crane in excess of its rated capacity.
Therefore, this DC 'mproved the operational safety of the Turbine Building crane. Tho crane and modifications made to it in to way affect nuclear safety.
DC 83-047 TITLE:
Augmented Off Gas A0G Recycle Elimination DESCRIPTION: The purpose of this Design Change was to perform process modifications to the Recombiner Train "B" in order to eliminate an off gas ignition source. This was accomplished by eliminating the recycle loop and replacing this dilution source with steam taken directly from the main steam supply.
SAFETY ANALYSIS:
This modification did not interface with any safe shutdown systems and was considered ron-essential. This modification enhanced the overall safety of the plant by reducing system maintenance and associated personnel radiation exposure, and improved system performance by reducing radioactive gaseous releases, and the likelihood of off gas ignitions.
p_C 85-023 TITLE:
Diesel Ger.erator No. 2 (DG2) Transformer Addition DESCRIPTION: The purpose of this DC was to provide a more reliable source of AC power for the Diesel Generator No. 2 (DG2)
Exhaust Fan 1-EF-DB-1B, DG2 Supply Fan 1-HV-DG-1B and DG2 Supply Fan 1-HV-DG-lD.
These fans were previously powered from MCC-DG2.
In a 10CFR50, Appendix 'R' scenario, a common fire could have disabled both feeders to MCC-DG1 and MCC-DG2.
This would have rendered DG1 and DG2 inoperative due to loss of cooling air.
Therefore, to eliminate this possibility, this DC installed a 250 k'!A transformer in the DG2 Room. The transformer taps off of the 4160 VAC Bus and the secondary of the transformer powers MCC-DG2.
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SAFETY ANALYSIS:
The addition of the 250 kVA transformer with - the associated protection equipment.will ensure that the Supply / Exhaust Fans will have reliable 480 volt power.
This DC meets the proposed' requirements of CNS Volume III, Response to 10CFR50 Appendix "R".
The current diesel generator isolation capability (DC 86-83, Diesel Generator Cable Isolation Switches, DG1 and DG2) along with the local control capability and the protected 125 VDC power feeds enables operation of the diesel generator completely independent of the fire areas requiring alternative shutdown capability, thereby increasing the overall safety of the plant.
DC 85-61 TITLE:
Training Center Construction DESCRIPTION:
The purpose of this DC was to construct a Training Center that will house the plant simulator and Training Department. The Training Center can also provide additional space for the Emergency Operating Facility (EOF).
The existing EOF and Training Center were connected by a hardened corridor.
SAFETY ANALYSIS:
This DC involved the construction of a Training Center outside of the protected area.
This DC did not affect any safety-related equipment nor did it degrade the overall safety of the station, and therefore, is considered non-essential.
The addition of the Training Center 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 no safety systems were affected and no work was performed in the area of a safety system.
DC 85-102 TITLE:
Laundry Room FP Sprinkler - Detection System DESCRIPTION:
This DC implemented an additional wet pipe Automatic Fire Protection Sprinkler System, with a flow detection device to the new laundry room.
The laundry room was moved to the 934'-0" elevation of the Radwaste Building and an additional Fire Protection System was required as a result of that move.
SAFETY ANALYSIS:
This DC provided a means of detecting and extinguishing a fire in the new laundry room. This DC did not affect the safety function of the Fire Protection (FP) System or any other plant systems iuportant to safety.
This system addition provides increased personnel and equipment safety in the new laundry room.
s
%! ^ itJ DC 86-034'A A
' TITLE:
'ATWS/SLC
. DESCRIPTION: This DC converted the existing Standby Liquid Control (SLC) System from ' a single pump - operation to af two '(2) pump operation to implement the ATWS rule documented in 10CFR50.62.' Due to increased flow, the : existing suction line did. not provide Net. Positive
' Suction Head. (NPSH) requirements for; two. pump operation.
Therefore,. separate suction lines from' the storage and test tanks were provided to each pump. Separate keylock on-off switches were installed for each pump.
Each switch will start its. respective pump, fire one explosive..(squib) valve in the SLC disc.harge piping and close one of the two Reactor Water Cleanup (RWCU) isolation'.
valves.
SAFETY ANALYSIS:
This DC provided two pump operation of the Standby Liquid Control System.and' did not degrade the function of any safety-related 4
. system.
.In addition the. installation of separate suction-lines to. each SLC pump will avoid' reinforcing pulsations and. ensure adequate NPSH thereby improving the operability of the SLC System.
i With the implementation of this DC the probability of' occurrence or. the ' consequences of. an accident or malfunction of' equipment important to safety as previously evaluated was not increased because all previous design features of.the SLC system were-
.. maintained or L improved upon.
Refer to Cooper Nuclear Station; Licensing' Amendment No. 123 dated July 5,1988.
DC 86-034B.
TITLE:
. Anticipated Transient Uithout Scram DESCRIPTION: This DC authorized the Alternate Rod Insertion -(ARI) and Recirculation Pump Trip (RPT) modifications as required by the NRC ATWS rule, 10CFR50.62. ARI provided a means.of reactor shutdown which is as~ diverse as practical and independent from the Reactor.
Protection System (RPS).
It is common to the normal shutdown components in the Hydraulic Control Units, control rod drives, control rods, and associated hydraulic system. components.
ARI valves were installed near the existing Hydraulic Control Units.
These valves provide an alternate method of depressurizing the scram valve pilot air header allowing the scram insert and discharge valves to open and the Scram Discharge Volume vent and' drain valves to close. Although the RPT met the criteria of the ATWS rule, it was modified to minimize the potential for inadvertent trips and to increase the level of redundancy associated with the existing RPT. The modifications consisted of I
adding a redundant trip coil and changing the trip logic to a 2 out of 2 logic that will trip both recirculation pumps.
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r SAFETY ANALYSIS:
This DC provided modifications that will reduce the risk from Anticipated Transient Without Scram (ATWS) Events.
These modifications are required by 10CFR50.62.
The modifications include Alternate Rod Insertion (ARI) and Recirculation Pump Trip (RPT). The work performed under this DC did 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 reducing the risk from ATWS events. This DC did not increase the possibility of an accident occurrence nor does it decrease the safety margin as defined in the Technical Specification.
The conceptual design of the ATWS modifications-1 and NRC acceptance of this conceptual design is described in General Electric Company Report No. NEDE-31096-P-A, " Anticipated Transients Without Scram -Response to NRC ATWS Rule,10CFR50.62".
General Electric Company Report No. 31113-P,
" Cooper ATWS Compliance: Alternate Rod Insertion System and Recirculation Pump Trip Modifications" describes plant specific modifications at Cooper Nuclear Station.
NRC Letter, W.
0.-Long, NRC to G.
A.
Trevors, NPPD, " Safety Evaluation Report on Cooper Nuclear Station Compliance with ATWS Rule 10CFR50.62 Relating to ARI and RPT Systems,...," dated September 23, 1987, provided NRC acceptance of this design.
DC 86-038 TI'rLE:
RG 1.97 Primary Containment Radiation Monitors DESCRIPTION: The purpose of this DC was to install environmentally qualified cables and connectors to the Victoreen High Range Containment Radiation Detectors. It ensures the operability of the Victoreen High Range Containment Radiation Monitoring System during a Design Basis Accident (DBA) by completely sealing the in-containment portion of the interconnecting cables.
SAFETY ANALYSIS:
This modification increases the Victoreen High Range Containment Monitoring System (VHRCMS) reliability by providing EQ cable and connectors and ensuring operability during a Design Basis Accident (DBA).
The implementation of this DC did not degrade plant personnel, equipment, and/or nuclear safety.
This DC did not present an unreviewed safety question nor did it change the function or operation of any system.
DC 86-040 TITLE:
Torus to Reactor Building Vacuum Breaker Isolation DESCRIPTION: This DC upgraded the Reactor Building to Torus Vacuum Breaker Valves PC-243 AV and PC-244 AV by adding spring to fail operators,
and upgraded the electrical configuration to NRC Reg. Guide 1.75. - _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _. _ _ _ _. - _ _ _ _ _ _ _.
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' SAFETY ANALYSIS:
- By providing qualified environmental, seismic, and hydrodynamic parts along with electrical separation this DC upgraded the Reactor Building to Torus Vacuum Breakers to bring them into conformance with the requirements for equipment regarded as safety-related.
The dual safety function was improved by replacing'the existing operators with a " spring to open" air operators, and installing 1983; and separate electrical new air tubing per ANSI B31.1 systems. Nuclear safety was increased by meeting single failure criteria of IEEE 279 and separation criteria of NUREG 1.75.
DC 86-060 TITLE:
Modification of Control Room HVAC Makeup Air System DESCRIPTION: This Design Change replaced the air intake damper of the Control Room Habitability System (CRHS) with a leak-tight butterfly valve to eliminate infiltration through the air intake when it is closed.
It also replaced the damper in the toilet / kitchen exhaust. duct with an 8-in. leak-tight butterfly valve to control air leakage through.
this-exhaust duct during the time that the filter booster fan is running.
And it also' sealed off the present 78-in. by 24-in, exhaust duct and changed the function of the existing exhaust 4
Fan 1-BF-C-1B from exhausting air and recirculating air to recirculating air only.
SAFETY ANALYSIS:
This Design Change was performed while the reactor was in cold shutdown with no fuel handling in progress.
The modifications performed under this Design Change did not in any way degrade Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. Implementation of these modifications assures the operability and reliability of the Control Room Habitability System (CRHS) by eliminating air leakage past the damper in the outdoor air intake and controlling air leakage from the Control Room / Cable Spreading Room envelope during all CRHS operating conditions.
DC 86-073 TITLE:
Primary Containment PT-512 Replacement l
DESCRIPTION: This DC authorized the replacement of pressure Transmitter PC-PT-512A and PC-PT-512B. These transmitters are used to measure drywell pressure.
In order to meet NUREG 0737 requirements for l
Control Room Design Review (CRDR) the new transmitters measure the I
drywell pressure in psig. The new transmitters were also required to meet separation criteria in IEEE-384 and the commitments set forth in the Districts response to Reg. Guide 1.97.
Divisional power supplies were used on the two transmitters.
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SAFETY ANALYSIS:
This DC involved. work on safety-related components.
This modification did not degrade the safety of Cooper. Nuclear Station (CNS) with respect to personnel, equipment, or nuclear safety.
The drywell pressure Transmitters PC-PT-512A and PC-PT-512B are intended to provide post accident monitoring capability.
These changes will increase the reliability of the drywell pressure instrumentation during and after an abnormal event or accident.
DC 86-083
. TITLE:
Isolation Switches for DG1 and DG2 Cable Isolation DESCRIPTION:
The purpose of this DC was to provide the capability to isolate control cables which run between the Diesel Generators Rooms 1 and 2 and the Switchgear Rooms for the 4160V Buses 1A and 1B.
These changes were determined to be necessary following various Appendix R evaluations which identified potential fire-induced spurious operations. See the District's response to Appendix "R",
Volume III. The isolation was accomplished by installing four new isolation switches on the Diesel Control Panels and the doors for Switchgear Breakers EG1 and EG2.
-Operation of the isolation switches-is annunciated in the Control Room.
SAFETY ANALYSIS:
This Design Change involved the addition of isolation switches in the DG system circuits. The addition of the new isolation switches 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 DG System remained unchanged. The possibility of spurious malfunction of the DG system circuits were mitigated, with the completion of this Design Change.
DC 86-085 TITLE:
Condensate Filter Demineralized DESCRIPTION:
This Design Change was written for the installation of an 8" bypass line in the Condensate Filter Demineralized System. This 8" bypass line, which contains a manual isolation valve and a control valve controlled by a differential pressure controller, will allow up to one (1). filter demineralized unit's flow to be bypassed while maintaining a constant system differential pressure.
The instrument air system is utilized for the air supply to the differential pressure controller and the valve air operator.
SAFETY ANALYSIS:
No systems important to nuclear safety were affected by this DC.
The implementation of this DC did 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 minimize the pressure fluctuations of the filter demineralizers during _ - _ - _ - _ _ _ _ _ - _ _ _ _ _ - _ _ - _ - _ _ _ _ _ -
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operation.- As a result, the' operating efficiency of the filter demineralizers-and-the life of the vessels' precoat and septums were increased.
DC 86-088 Amendment 1 TITLE:
Installation of Area Radiation Monitors in MPF DESCRIPTION: The purpose of this Design Change Amendment.was to install area radiation monitors in the Multi-Purpose Facility (MPF). Building to facilitate safe storage and processing of Low Level Radioactive Waste without radiation exposure exceeding that allowed by.10CFR20
and 10CFR190.
Radiation exposure to personnel will be reduced.
with the addition of the area radiation monitors because individuals will be warned of higher than normal radiation levels.
SAFETY ANALYSIS:
This modification, considered non-essential, had no effect on' overall plant-safety nor'did it interface with any safe shutdown systems.
This modification enhanced the safety of the plant by ensuring that-personnel will be warned of higher than. normal radiation levels.
DC 86-155 TITLE:
Undervoltage Relay Replacement
. DESCRIPTION: This Design Change ' involved the' replacement. level protection General Electric (G.E.) ' Type IAV54E _ second ~
four existing of undervoltage relays for the Critical 4160V Busses with Brown-Bovari' Type ITE-27N undervoltage relays which have a definite time delay.
The G.E. IAV54E relays were not ' specifically designed for ' the application in:which they were used.
Depending on the profile curve of voltage over time, variable' response time results could be outside the required time delay specifications. The Replacement' Relays are definite time relays (Brown-Boveri Type ITE-27N Undervoltage Relays) specifically designed' for this type of application. :The Brown-Boveri ITE-27N relays operate and comply with the Technical Specification of'7.5 0.8 seconds.
Refer to Cooper Nuclear Station Licensing Amendment No. 124 dated August 3, 1988.
The Agastat Timers, 27X7/lF and 27X7/lG, will remain connected in series with the new 2nd level Undervoltage Relays.
However, the Agastat Timers were reset for 5 seconds i 0.5.
SAFETY ANALYSIS:
The replacement of the four G.E. Type IAV54E undervoltage relays with BBC Type ITE-27N undervoltage relays that have a definite time delay required for second level undervoltage protect. ion will assure proper critical 4160V switchgear undervoltage protection.
The installation did not violate any Technical Specification nor reduce the margin of safety as specified in CNS Technical Specifications, Table 3.2.B, applicable to the 4160V critical switchgear bus undervoltage relays.
Refer to Cooper Nuclear Station Licensing Amendment No. 124 dated August 3, 1988.
The probability of or - _ _ _ _ _ _
r consequences of an accident or malfunction of equipment important to safety is reduced because the new BBC undervoltage relays were specifically designed to perform this safety function (second level undervoltage protection).
No possibility for an accident or l
malfunction of a different type than previously evaluated in the USAR exists.
DC 87-015K TITLE:
Multi-Channel Gaitronics DESCRIPTION: This Design Change covered the upgrade of the Industrial Communication (Gai-Tronics) System from a two channel (1 party-line and 1 page channel) to a six channel (5 party-lines and 1 page channel) System for the main structures only, i.e.
Turbine Generator, Reactor, and Control Buildings.
The new six channel system allows for a greater number of separate, yet simultaneous conversations. The upgrade entailed replacing existing desk-top, side desk, and wall-mounted two channel Plant stations with new six channel stations, installing new 16 conductor cabling, and revising some conduit routings.
In addition, new stations were added in the Condensate Pump Room, Diesel Generator Rooms, Reactor Feed Pump Room, South and North Heater Bay door areas, Hotwell, SAS/CAS Consoles, New Condensate Storage Tank / Fire Tank Area, and near the Off-Gas Sample Station.
System zones were established to provide separately fused, independent circuits to facilitate maintenance and assure high system reliability.
. SAFETY ANALYSIS:
This DC involved work on non-essential systems and did not degrade CNS with respect to personnel, equipment or nuclear safety, and was considered non-essential. The Industrial Communication System Upgrade did not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR because no Plant safety system's function, including the Industrial Communication System's function, were changed or altered in any manner. Upgrading the existing system to a multi-party line system did not increase the probability of occurrence or the consequence of an accident because the Industrial Communication System is completely independent from all other Plant systems. The system does not perform any safety function nor support any safety-related equipment.
DC 87-017 TITLE:
Permanent Installation of Radwaste Dewatering System DESCRIPTION: The purpose of this Design Change was to permanently install radwaste resin dewatering equipment.
The new system processes powdered and bead type ion exchange resins and other filter rr.edia by removing excess water from the resins.
The entire operation of the Resin Dewatering System is by remote control (with the exception of replacing and removing the fillhead assembly and __
', L * '
l l
e installing the liner cap), which provides ample protection from contamination or mechanical hazards to personnel.
SAFETY ANALYSIS:
This installation did not present a safety hazard to Cooper Nuclear Station operations.
This modification did not affect any vital areas or safety-related equipment, and as such, does not present a threat to nuclear safety..This modification did not present a reduction of personnel safety because adequate shielding is available to attenuate radiation effects.
The entire operation of the system is by remote control (with the exception of placing and removing the fillhead from the liner and installing the liner cap) which provides protection from contamination or mechanical hazards to personnel. The existing radwaste processing capability was not degraded by this installation.
Equipment safety is improved above the previously existing arrangement because much of the temporary electrical and mechanical installation was replaced by more durable, permanent installations (i.e. hard pipe versus, flexible hose, electrical conduit versus open cable, etc.).
DC 87-036 and Amendment 1 TITLE:
Containment Hydrogen /0xygen Monitoring Systems DESCRIPTION: The purpose of this Design Change was to replace existing hydrogen / oxygen (H /0 ) analyzers with H /02 analyzers qualified 2 2 2
to IEEE 323-1974 and Regulatory Guide 1.100 requirements, meeting the District's Regulatory Guide 1.97 Commitment.
The new H /02 2
H /02 analyzer panels, monitoring system consists of redundant 2
remote control panels, and recording equipment. The H /02 analyzer 2
panels are mounted in the Reactor Building, Elevation 976'-0".
The remote panels are mounted in the Cable Spreading Room.
All alarm annunciation and associated recording equipment are mounted in the Control Room, on Vertical Boards P1, P2 and H.
All equipment was mounted using seismically verified mounting hardware and methods. Also, the new H /02 systems provides Plant Management 2
Information System (PMIS) signal output to MUX Cabinets 9-82 (Div. I) and 9-83 (Div. II).
SAFETY ANALYSIS:
Implementation of this DC did not, in any way, degrade the safety of Cooper Nuclear Station with respect to personnel, equip;nent, or nuclear safety. It was the intent of this DC to upgrade plant safety by installing qualified equipment to ensure compliance with Reg. Guide 1.97, Rev. 2 monitoring requirements. This DC did not increase the possibility of an accident occurrence, nor did it decrease the safety margin as defined in the basis of any Technical Specifications. System operability and reliability are increased as compared to the original design. The new systems are qualified per the above-mentioned Reg. Guide 1.97 requirements, which provides assurance that operation will be maintained throughout a seismic event or design basis accident. -
e DC 87-037 TITLE:
Torus Level and Pressure Transmitters DESCRIPTION: This DC replaced existing transmitters (PC-DPT-3Al and PC-DPT-3B2) and added new transmitters for wide range torus pressure (PC-PY-30A and PC-PT-30B) in' order to meet the requirements of NUREG 0737 Supplement 1 as described'in Regulatory Guide 1.97 (R.G. 1.97).
The original torus level transmitters, which were manufactured by Barton Instruments, were' not environmentally and seismically qualified.
The replacement transmitters meet the environmental and seismic requirements described in R.G.1.97.
The'new wide range torus pressure transmitters were installed because no existing transmitters met the range requirement (-5 to 70 psig) of R.G. 1. 97. The power supplies and recorders required for wide range torus pressure already existed.
The new wide range torus pressure signals were also routed to Plant Management Information System (PMIS).
SAFETY ANALYSIS:
The installation of the seismically and environmentally qualified transmitters ensures that this equipment will' perform its safety function during a design basis event.
This increases the reliability ' of the components (compared to previously existing) to operate as intended during any accident condition, thereby increasing the safety margin of each system. The addition of the torus wide range pressure instrumentation enhances operator action following an accident, thereby increasing the safe operation of the plant.
DC 87-043 TITLE:
Replacement of NBI Pressure Switches DESCRIPTION: The purpose of this Design Change was to replace NBI-PS-51A, B,.
C, D and NBI-PS-52A, C with switches qualified to NUREG 0588 Category I, IEEE 323-1974 and 10CFR50.49 requirements.
The Barksdale Switches (located in a harsh environment) were replaced with qualified Static-O-Ring switches.
All set points were retained. Switches NBI-PS-51A,B,C,D were replaced on a one-for-one basis while NBI-PS-52A, and 52C were replaced with two switches each to account for the dual element function of those switches.
SAFETY ANALYSIS:
The implementation of this Design Change did not reduce the margin of safety since all safety functions, set points, and system operational parameters were retained. The overall safety of the NBI system is improved since the switches that were installed have been tested and proven capable of performing their safety function in the event of a design basis accident. _ - _ - _ _ _ _ _.
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DC 87-069 1
TITLE:
Gamma TIP Installation DESCRIPTION: The purpose of this DC was to provide a more accurate core power distribution evaluation. The original thermal neutron-sensitive Traversing Incore Probe (TIP) were extremely sensitive to the instrument tube (and hence instrument) location in the water gap
'between fuel bundles.
This sensitivity to location allowed for.
a larger uncertainty factor input, to the process computer power distribution, and approach-to-limits calculations. -By replacing.
the thermal neutron TIPS with gamma-sensitive TIPS that are not as sensitive to water gap location, the core power distribution calculation is more accurate..The gamma TIP system was a direct changeout of the. neutron TIP system to minimize existing replacement procedures.
SAFETY ANALYSIS:
The TIP detectors are non-essential and provide no safety function.
The gamma TIP changeout did not impact any safety system.
The core performance monitoring systems were improved by providing a more accurate power distribution measurement.
The improved accuracy is also an input to the core thermal limits evaluation.
The TIP system provides calibration for the LPRM system to account for depletion of the U-236 over time. The calibrated LPRM signals are used as input to the APRM safety system.
DC 87-073.
TITLE:
Replacement of 125 VDC Station Batteries, Racks, and Chargers.
DESCRIPTION: The purpose of this Design Change was to replace the 125V Batteries lA and 1B, including the battery racks, the 125V Battery Chargers lA and 1B, the 480V, 60A fusible disconnects in MCCs LX and TX, and the remote indicating ammeters for Chargers lA and 1B (located on Vertical Board C in the Control Room).
A third charger, designated 1C, able to operate in parallel with either Charger lA or 1B, was also added.
Manual Transfer Switches and fusible disconnects were installed so that Charger 1C can be used l
in parallel with either Charger lA or 1B. Charger 1C, under normal operating conditions, will not be in service.
The batteries in the 125V system were replaced by lead-calcium batteries which have a higher ampere-hour rating.
The battery chargers for the 125V system were also increased to a higher amp rating.
Due to the increased size of the replacement battery cells, it was necessary to replace and rearrange the battery racks.
The existing DB-25 breakers, which are associated with the DC output of Chargers lA l
and 1B, had their coil assemblies changed from 200 ampere coils to 300 ampere coils.
The 300 ampere coils provide better coordination with the feeder breakers.
The existing DB 50 breakers, which are the 125V Battery feeders to the 125V Switchgear Buses lA and 1B, had their coil assemblies changed from 800 ampere coils to 1600 ampere coils.
The new 1600 ampere coils provide '
p c., 3 e
=,
1 better ' coordination E with the system feeder. breakers; For
. additional information, refer to Cooper Nuclear Station Licensing; Amendment No. 122 dated June 16, 1988.
SAFETY EANALYSIS:
The existing safety function of the-125V DC Power System remains unchanged.-
No additional _ safety concerns beyond those already evaluated in the USAR result from this design change. No Technical Specifications were violated. ' The margin of safety. associated with the?l25V DC Power System was maintained. Refer to Cooper Nuclear Station. Licensing Amendment No. 122: dated June 16', 1988. System
~
performance was improved. The original batteries were. nearing the end of their design life' and tested at less than 90 percent of the.
manufacturer's. rating.
The new batteries have a : rating of' 1800 ampere. hours, as compared to~ the. original.125V batteries'-
i rating of 1672 ampere hours. Also, with the addition of the third-battery Charger.1C, reliability' is also increased because it will.
enable operators to transfer to a live bus in the event one train:
.of ACl power is disabled.
DC 87-088 and Amendment 1 TITLE:
Add Check Valves to the Drywell Sump Isolation Valve' Accumulators -
DESCRIPTION: The purpose of this Design Change was to install four (4) 1/4" check valves in the Drywell Radioactive Equipment and Floor Drain air supply systems.
The primary' purpose of these check valves
.is to provide an essential, Seismic Class IS air supply for closing
' four ' primary, containment isolation valves upon - receipt of an isolation signal should a loss of air event occur. The new check.
valve, existing accumulators, air supply tubing, and. solenoid valves have all been, qualified to essential, Seismic'IS requirements resulting in a system which is equivalent to a spring to-close system.
SAFETY ANALYSIS:
With the implementation of this DC, primary containment is enhanced because upgrading the closing mechanisms for the Drywell Sump.
Primary Containment Isolation Valves to a Seismic IS configuration-increases the reliability of the system to operate during certain postulated scenarios.
This DC did not affect any other safety system nor was the system performance. changed.
However, this system has been improved by providing a fail closed valving system for the Drywell Sump Isolation Valve Accumulators.
DC~87-096 TITLE:
Reactor Water Level Instrumentation DESCRIPTION: The previously installed reactor water level instrumentation did not meet the range or qualification requirements of Regulatory '
Guide 1.97. To meet the requirements of Regulatory Guide 1.97 for water level instrumentation, this Design Change modified two existing ranges and installed a third range. The reactor vessel <
i 3
.g ce;,
-a water level instrumentation provides operators with indication of-reactor water level to verify that the core is adequately covered by the coolant inventory inside the reactor vessel. Vessel level is indicated locally in the Reactor Building and remotely in the Control Roon..
Two existing qualified transmitters (NBI-LT-91A, B) and one new environmentally qualified transmitter on Rack 25-51 are used for the fuel zone range water level. The cold condensing chamber provided the reference for these transmitters. Three new environmentally qualified transmitters on Racks 25-5 and 25-6 were installed for wide range water level. The cold condensing chamber provided the reference for these transmitters. The existing wide' range level transmitters (NBI-LlT-59A, B) and their associated 1
signal conditions (NBI-CU-59A, B) were. removed by this DC.
.The new wide range transmitters will perform the function that these transmitters were performing.
SAFETY
' ANALYSIS:
The increase' in the range of the fuel zone Rosemount level instruments and the addition of the steam nozzle range instruments
(
cannot create any new or different kind of accident. This change I
increases the range, by providing - accurate level measurement capability from the core support plate to above the centerline of.
the main steamlines. These level instruments provide indication j
of water level in the vessel, but do not provide any safe shutdown 1
or ESF input signals. The range increase provides a better water level assessment capability and is in accordance with NRC guidance provided in Regulatory Guide 1.97.
The increase.in the range of i
NBI-LI-91 A and B and the addition of the steam nozzle range cannot create any new or different kind of accident, since they are used for assessment purposes only.
Also, _since environmentally qualified equipment is used and a third indication is being-installed in the fuel zone to resolve any ambiguity in the two redundant indications, this change decreased the probability of false level indication contributing to any new or different kind of accident.
DC 87-113 TITLE:
Reactor Water Level Reference Leg DESCRIPTION: The purpose of this Design Change was to bring CNS into compliance with the District's commitments in response to Generic Letter 84-23.
This was accomplished by providing a means of injecting water into the reactor vessel water level cold reference
')
legs (no temperature compensation) such that the effects of l
flashing and boil-off of the reference leg water will be precluded.
l The core spray pumps will be utilized to deliver suppression pool I
water through a valve station and into the reference legs at the l
instrument racks where the reference legs terminate. The injection water will refill the reference legs up to condensing Chambers 3A and 3B, ultimately discharging into the reactor vessel.
The reference leg injection system is completely manual. No automatic signal opens or closes the solenoid valve. It is not interlocked to any other system, nor does solenoid actuation act as an input _ _ _ - _ _ _ -
3 ? '>-
C 1
to any device-other than the solenoid valve position indicating lights located.in the control room. Actuation and termination is achieved solely by the operator placing the solenoid valve handswitch located in Control Room Panel 9-3 in the "0 PEN" and "CLOSE" position.
This simplicity in operation and controls' results in greater system reliability.
This modification implements the Districts commitments made in response to Generic Letter 84-23 and for closing out NUREG-0737, Item II.F.2.
SAFETY ANALYSIS:
The addition of the Reactor Water Reference Leg will lessen the probability of inaccurate vessel level indication occurring under certain accident scenarios.
Under these conditions this change is an improvement on the original design since it will provide a more accurate indication of the reactor water level for verification that the core is adequately covered by the coolant inventory present in the reactor vessel following a design basis accident. This. change cannot create any new or different kind of accident, since it improved the existing water level measurement system, but did not change the function. The margin of safety as defined in the Technical Specifications is not reduced since the modification only enhanced water level assessment capability following certain DBAs.
DC 87-133 TITLE:
Replacement of Overcurrent Relays 51/IFE and 51/ IGE DESCRIPTION: This DC involved the replacement of the existing General Electric Overcurrent Relays 51/IFE and 51/ IGE with Basler Overpower Relays Type bel-32. The new relays are single-phase directional overpower relays. The directional overpower relays measure electrical power output and use the direct relationship of horsepower to watts to initiate non-critical load shedding before stall-horsepower has been reached on the Diesel Generators.
SAFETY ANALYSIS:
The installation of the new relays provide increased system stability and improve the safety of the plant. The new overpower relays will shed non-critical loads from the emergency diesel generators before engine stall horsepower is reached, thereby assuring the 4160V AC power is provided to Critical Buses IF and 1G to run critical plant equipment. The margin of safety is not reduced nor is the possibility of an accident or malfunction created or increased with the implementation of this design change.
DC 87-152 and Amendment 1 TITLE:
Sequential Starting of RHR and CS Pumps from Start-Up Transformer DESCRIPTION: This Design Change authorized the modification of the RHR and Core Spray Pump start logic to initiate a sequential start of the pumps when powered from the Start-up Transformer.
This Design Change involved logic changes to Relay Panels 9-32 and 9-33 in the __-_____________________
.1
}([*'Sf ji Auxiliary Relays Room.
These changes allow the ; ECCS - pumps : to
- sequentially start. on the Start-.up - Transformer using. the same k
, time-delay relays and-start relays as the Diesel-Cenerator and 1 Emergency Transformer. These changes did not affect the Electrical
. Separation criteria or seismic, qualification' of.the relay panels. -
x SAFETY-
.' ANALYSIS:
The implementation of this DC did not degrade the safety of Cooper Nuclear Station with respect to. personnel, equipment, or nuclear safety. tThis'DC'did not increase the possibility of an accident-v occurrence.
This modification increased: the. safety marginiand '
improved"the voltage control for. the Auxiliary Power Distribution l
System.
This Design Change will. eliminate the potential large voltage dip on the Critical 4160 Volt. Buses during the block start of.the-six ECCS pumps'on a:LOCA initiation'and allow thei161.kV~'
Line to operate at a wider voltage tolerance, thereby reducing the possibility. of the plant' entering' a ;LCO due to' low volta'ge on the-161 kV line or the unavailability of the 345/161 kV-transformer.-
DC 87-170 bm TITLE:
Suppression Chamber Cooling Throttle MOV Modifications
' DESCRIPTION:. Design Change 87-170 authorized the replacement of the 150'ft.-
lb., 3400 RPM motors on RHR-MOV-M034A(B) with 200 ft.-lb.,1700 RPM motors.
The first-reduction gear set
.of the Limitorque SMB-4 operators was changed. This resulted in a change in the overall gear ratio from 77:1 to 72.62:1. This modification is intended to eliminate the problem'of motor pinion key shearing.
The stroke time of the MOV assembly was increased from 24 seconds to 39 seconds.
SAFETY ANALYSIS:
This Design Change did not increase the probability of occurrence
.or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
The previously evaluated accident under consideration here is the Loss-of-Coolant Design Basis Accident discussed in Section XIV of the USAR. The increased valve stroke time of 39 seconds will still result in the suppression chamber cooling throttle valves being closed prior to the start of full LPCI inj ection.
This modification did not create any new mode of operation or affect any operating limits or setpoints.
DC 87-176 TITLE:
Secondary Containment Ventilation Isolation System Modification DESCRIPTION: This Design Change provided the modifications necessary to enhance the seismic qualification of the air lines, tube trays, and duct supports for the air system that provides instrument air to operate the isolation valves on the Secondary Containment Ventilation Isolation System. It also replaced a switching valve on each air b ____--______
4 accumulator tank with a new seismically qualified check valve to prevent a loss of pressure due to failure of instrument air.
SAFETY ANALYSIS:
The modifications performed under this DC did not in any way degrade Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety.
Implementation of these modifications further assures that the Secondary Containment Ventilation Isolation System will maintain integrity and will function after a design basis seismic event. Since the replacement of the air accumulator tank control valves were coordinated with the plant in a cold shutdown condition with no fuel manipulations in progress, plant safety was not affected by the secondary containment ventilation system being rendered inoperable while the modifications were made.
DC 88-154 TITLE:
CRD Test Connections DESCRIPTION: The purpose of this Design Change was to install four 1/2" test connections in the 3/4" Control Rod Drive Hydraulic Piping System.
The test connection routing consists of a straight piping segment with two socket welded 1/2" valves which expand to a 3/4" threaded end for test equipment connection. The connections were installed to perform the Primary Containment Local Leak Rate Test on the following Isolation Check Valves:
CRD-CV-13CV and CRD-CV-15CV outside containment and CRD-CV-14CV and CRD-CV-16CV inside containment.
SAFETY ANALYSIS:
This modification improved the safety of the plant by ensuring that the Isolation Check Valves CRD-CV-13CV,14CV,15CV, and 16CV have a local leak rate test performed on them.
Local Leak Rate Testing of these valves were required by the NRC W. O. Long letter to G. A. Trevors dated January 13, 1988.
This modification involved no other safety function and did not degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety.
DC 88-239 TITLE:
Security Door Isolation DESCRIPTION: This Design change permanently barricaded vital doors H-114 and R-115 by using 1/2" steel plate to cover the door frames outside the room or building.
Non-Vital Door H-307 was also barricaded from inside the Computer Room by the use of a deadbolt/ lock arrangement. These modifications eliminated unauthorized access through the doors and into the Reactor Building, Control Building, and Computer Room and meet the new Safeguards Security Plan that went into effect on November 1, 1988. _ _ _ _ _
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~
s SAFETY ANALYSIS:
The implementation of this Design Change did not degrade plant personnel safety, equipment safety, or nuclear safety during or following the modification. By barricading these doors shut, CNS' still maintains two (2) exits out of each area as designated by the.1988 edition of NFFA 101 Life Safety Code. Reactor Building Door R-ll5 is considered to be an essential door for the purpose of maintaining secondary containment. However, this Design Change did not alter its ability to maintain secondary containment.
ESC 88-276 TITLE:
Zircaloy-2 Fuel Bundle Spacers and Channels DESCRIPTION: The purpose of this Equipment Specification Change'(ESC) was to provide technical justification' and authorization in using fuel bundle spacers and channels fabricated from an improved Zircaloy-2 material.
The applicability of this new material is for all present and future fuel bundle types to be used at Cooper Nuclear Station.
SAFETY ANALYSIS:
This change represents a replacement of an existing piece of equipment; the mechanical properties of Zircaloy-2 and Zircaloy-4 are nearly identical (the slight reduction in tensil strength is not relevant to the application). There were no required changes to P & ids or-one-line electrical drawings, there was no impact-to other plant systems, the design basis of associated components was unaltered, and there is no cumulative effect to related systems. Therefore, the margin of safety was not reduced nor was the possibility of an accident or malfunction created or increased with the implementation of this ESC.
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III.
PERSONNEL AND MAN-REM EXPOSURE BY WORK AND JOB FUNCTION
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JL d '.'r PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION Number of Personnel Total Man-Rem
(> 100 mrem) l Station Utility Contractor Station Utility Contracto:l Work and Job Function Employees Employees
& Others Employees Employees
& Others REACTOR OPERATIONS & SURV.
Maintenance Personnel 9
0 8
0.579 0.000 0.664 Operating Personnel 50 0
0 17.666 0.000 0.000 llealth P.tysics Personnel 26 0
11 5,941 0.000 1.219 Supervisory Personnel 4
2 1
0.659 0.419 0.112 Engineering Personnel 17 17 19 4.599 2.642 3.629 ROUTINE MAINTENANCE Maintenance Personnel 72 3
221 61.557 1.901 93.652 Operating Personnel 1
0 0
0.001 0.000 0.000 Health Physics Personnel 22 0
13 15.493 0.000 5.648 Supervisory Personnel 1
2 0
0.098 0.120 0.000 Engineering Personnel 4
23 11 0.069 8.231 1.319 SPECIAL MAINTENANCE Maintenance Personnel 3
0 1
0.181 0.000 0.013 Operating Personnel 1
0 0
0.059 0.000 0.000 health Physics Personnel 2
0 1
0.086 0.000 0.015 Supervisory Personnel 0
0 0
0.000 0.000 0.000 Engineering Personnel 0
0 0
0.000 0.000 0.000 WASTE PROCESSING-1 0
0 0.001 0.000 0.000 Maintenance Personnel Operating Personnel 7
0 0
2.947 0.000 0.000 Health Physics Personnel 8
0 0
0.421 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 REFUELING Maintenance Personnel 0
0 0
0.000 0.000 0.000 Operating Personnel 34 0
0 1.490 0.000 0.000 Health Physics Personnel 1
0 1
0.042 0.000 0.046 Supervisory Personnel 2
0 0
0.103 0.000 0.000 Engineering Personnel 4
0 0
0.272 0.000 0.000 INSERVICE INSFECTION Maintenance Personnel 0
0 14 0.000 0.000 4.549 Operating Personnel 4
0 0
0.079 0.000 0.000 Health 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 1
0.000 0.000 0.035 TOTALS Maintenance Personnel 72 3
236 62.318 1.901 98.878 Operating Personnel 52 0
0 22.242 0.000 0.000 Health Physics Personnel 27 0
13 21.983 0.000 6.928 Supervisory Personnel 4
2 1
0.860 0.539 0.112 Engineering Personnel 17 26 20 4.940 10.873 4.983 GRAND TOTALS 172 31 270 112.343 13.313 110.901
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'[i GENERAL OFFICL h,.
Nebraska Public Power District
" " "LEso"a"Ssfa^ sat *** ***
S i
o, NLS8900094 February 28, 1989 U.S. Nuclear Regulatory Commission i
Document Control Desk Washington, DC 20555 Gentlemen
Subject:
Annual Operating Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 In accordance with Paragraph 6.5.1.C 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,
1988, through December 31, 1988.
We are enclosing one signed original for your use and, in accordance with 10 CFR 50.4, are transmitting one copy to the NRC Regional Of fice, 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, 12
. Trevors Division Manager Nuclear Support CAT /tja jw Attachment cc NRC Regional Office Region IV NRC Resident Inspector Cooper Nuclear Station Division Manager of Nuclear Operations w/l Enclosure Cooper Nuclear Station 1'7 IQ fh*Wl$$e1llllN OU$$TWUA5$!YANYN h 5 5 I?h ki
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