ML20090B697
| ML20090B697 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1991 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD920105, NUDOCS 9203040030 | |
| Download: ML20090B697 (29) | |
Text
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GENERAL OFFICE P O. BOX 499. COLUMBUS, NEBRASKA 68602-0499
-Nebraska Public Power District
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NSD920105 February 28, 1992 U.S.-Nuclear Regulatory Commission 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, hereby, submits the Cooper Nuclear Station Annual Operating Report for?the period of. January 1, 1991, through December 31, 1991.
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 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,
/-
G.
Horn Nuclear Power-Group manager GRH/tja:91an-rpt.ltr
. Attachment cc:
NRC Regional Office Region IV NRC Resident Inspector Cooper Nuclear Station
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COOPER NUCLEAR STATION BROWNVILLE, NEBRASKA ANNUAL OPERATING REPORT JANUARY 1, 1991 THROUGH DECEMBER 31, 1991 I
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TABLE OF_ CONTENTS.
SECTION PACE 1
I.
PERFORMANCE CHARACTERISTICS,
Puel Performance.
2 MSV and MSRV Failures and Challenges 3
I II.
FACILITY CHANGES,-TESTS, OR EXPERIMENTS REPORTABLE UNDER 100FR50.59 4
. Reportable Special Procedures /Special Test Procedures 5
112 Reportable-Design Changes Reportable Activities (Setpoint, Procedure Changes) 22 III.
PERSONNEL-AND MAN-REM EXPOSURE 25 1
26 By Work-and Job Function l'
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. PERFORMANCE-CHARACTERISTICS 1.
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l FUEL PERFORMANCE Cycle 14 operatian continued from January -1, 1991 through March 22, 1991.
Operation of the unit was intarrupted on March 23, 1991, via a manual scram, for
- repair associated with a feedwater check valve, The unit was restarted on March p
27, 1991 and continued operation until May 9,1991. The unit was shutdown on May 9, 1991 for repair of a Core Spray Valve. Startup of' the unit commenced on May
.10, 1991 and operation in Cycle 14 continued until October 4, 1991.
Cycle,14 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 CNS Technical Specifications. Comparisons of actual control rod densities to the control rod
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densities predicted by computer program calculations at various core average i
exposures, indicated no reactivity anomalies of 1% or greater.
During the period from October 4,1991 through December 14, 1991, the reactor was l
shutdown and the reactor vessel disassembled for the scheduled refueling and L
maintenance outage.
A core offload and reload was performed. which included replacement of 164 fuel assemblies. With the concurrence cf General Electric, i
it was decided that sipping for leaking fuel assemblies was not warranted due to extremely low off-gas activity.
Cycle 15 operation commenced with initial reactor startup on December 15, 1991 and 100% thermal power was initially achieved on December 26, 1991. The startup physics test program was completed on January 13, 1992, with notification of test completion to be submitted to the NRC under separate correspondence.
Cycle 15 operation continued through December 31, 1991.
Off-gas activity
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continued at essentially steady-stato levels with reactor coolant dose equivalent I-131 equilibrium values and off-gas release rates maintained well within the limits specified by CNS Technical Specifications. Comparisons of 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% or greater, L.
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i MSV AND MSRV FAILURES AND CHALLENGES (Ref.: NUREG-0737, Action Item II k.3.3)
There were no operational failures or challenges to the Main Safety Valves or Main Steam Safety Relief Vr.~.ves during the operational year-of 1991.
The _" challenges" made in accordance with Technical Specification 4.6.D.1, cyclic testing, were documented in LER 91 015.
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II.
FACILITY-CHANGES,- TESTS.- OR EXPERIMENTS REPORTABLE UNDER 10CFR50.59 l-l l
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REPORTABLE SPE'CIAL PROCEDURES / SPECIAL TEST PROCEDURES SP 86 004 TITLE:
PMIS Functional Check DESCRIPTION:
This Special Procedure (SP) provided specific instructions to operations personnel regarding verification checks during testing of the Plant Management Information System (PMIS) installation. This Special Procedure performed these checks to ensure that the PMIS/SPDS (Safety Parameter Display System) was completely functional for use by CNS Operations personnel.
SAFETY ANALYSIS:
This SP was intended only to functionally check the PMIS/SPDS operation. No systems were directly affected by this SP. The PMIS computer syst:m is part of the plant monitoring system only, and does not directly affect the performance or operation of any plant system.
This SP did not create an accident or malfunction of a different type, nor doct ease the margin of safety of CNS because it is only a plant monitoring system and does not directly affect the operation of any plant system including those important to safety.
SP 90-231 TITLE:
In-Service Testing of REC-CV-16CV DESCRIPTION:
The purpose of this Special Procedure (SP) was to test the operation of REC-CV-16CV.
ASME Section XI, requires all valves which are required to operate in order to perform their safety function be routinely tested.
This SP simulated a depressurization of the non-critical portion of the Reactor Equipment Cooling (REC) System by o'pening two temporary installed valves utstream of the REC-CV-16CV and verifying closure of REC-CV-16CV by observing a reduction in flow.
SAFETY ANALYSIS:
This SP did not increase the probability of an accident or i
malfunction of equipment important to safe ty previously l
evaluated in the USAR. This SP did not shut down the critical REC loops or jeopardize the operability of the essential portions of the REC System.
The SP was written to maintain the critical loops of REC operable.during the testing of REC-CV-16CV.
A partial REC System outage was required, but-only shut down the non-critical, non-essential loops of the REC.
The plant was in cold shutdown status (<212 *F) during the test with fuel pool cooling being the only non-critical load requiring support.
Additionally, Spent Fuei Pool Cooling could have been provided by Residual Heat Removal System, if necessary.
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SP 90-235 TITLE:
Post Accident Sampling System (PASS) Line Backflush Procedure DESCRIPTION:
The purpose of this _Special Procedure was to backflush the PASS to increase the flow rates by removing any potential-foreign material or other obstructions that could possibly cause a partial-flow restriction in this system.
SAFETY ANALYSIS:
The PASS has no safety function.
The only function of the PASS'is to allow collection of reactor coolant, torus water, and contalienent atmosphere during accident conditions, to assist operations personnel in evaluating _ the severity of nuclear accidents.
This procedure did not change the form, fit, or function of the PASS.
In addition, all PASS sample
. lines were _ isolated from reactor _ systems during - this procedure, and the pressure used to backflush was less than the rated capacity of the tubings and fittings in the PASS j.
Also, this procedure included steps to prevent discharged material from becoming airborne or discharged into the Reacto-l Building, and the flushing medium was non-toxic.
Therefore, performing this Special Procedure did not create an unroviewed safety question.
STP 87-008 TITLE:
Rod Worth Minimizer Operability DESCRIPTION:
This Special Test Procedure (STP) was designed to accomplish a
thorough evaluation of the Science Applications International Corporation (SAIC) installed improvements to the Rod Worth Minimizer (RWM).
The purpose of this STP was to establish -that the corrections made to the RWM program l
eliminated - previously identified - program deficiencies, and l
that no new significant problems remained af ter these changes L
were installed.
SAFETY ANALYSIS:
Performance of this STP involved movement of control rods (limited to one notch) while in the 75%. power region.
A Rod Withdrawal Error Analysis of this sort has been analyzed in the USAR (Section XIV-5.1.5) for the continuous withdrawal of the fully inserted maxfinum worth rod at. maximum drive speed.
This STP was bounded by the worst case condition in which the reactor is in its most reactive state with no xenon or-samarium present in the core.
Additionally, a limiting l:
control rod pattern surrounding this maximum worth rod was l_
imposed such that the withdrawal of the control rod would not l
breach thermal limits.
This STP did not change the plant facility or procedures as described in the USAR ' or the Technical Specifications. All safety aspects were reviewed, and there was no possibility of an accident or malfunction of equipment important to safety as a result of performing this test procedure.
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STP-87-021 Computer Point Installation / Data Collection TITLE:
RRMG Set 1
DESCRIPTION:
This Special' Test Procedure was performed to investigate certain minor undemanded speed changes previously observed in the Reactor Recirculation Motor Generator -(RRMG) Control System. Additionally, this STP temporarily modified each RRMG Set's control system to gather data with the intent to monitor and troubleshoot the control system.
SAFETY.
ANALYSIS:
.The installations performed as a part of this STP only affected the speed control circuit of the recirculation pumps, and did not alter the effects of a recirculation pump trip, i
nor did this STP alter any of -- the logic initiating a pump j
trip.
Therefore, by not affecting the results of recirculation pump trips, the probability or consequences of-an accident analyzed in the USAR or Technical Specifications was not increased, or the possibility of an un-analyzed l
accident created.
.l SIP 88-002 TITLE:
Kaman Interrupt Removal DESCRIPTION:
The purpose of this Special Test Procedure was to collect data j
on the. Elevated Release Point (ERP) Kaman high range high (HRH), and high range normal (HRN) radiation monitors during normal operating conditions.
This STP also performed tests that removed jumpers for the unused interrupts on the system and CPU boards and required the motJtors to be off-line, i
~ SAFETY LANALYSIS:
This Special Test Procedure did not degrade Cooper Nuclear
. Station with respect to personnel, equipment, or nuclear safety. This test required the ERP Kaman monitors to be of f-line but the alternate G.
E.
sampling system was put into service to sample off-gas radiation levels at the stack.
Technical Specifications allows for maintenance and testing of this equipment, and the monitors were returned to service with in their specified limits.
STP 88 -189 l
TITLE:
Trial Chemical Treatment Of Circulating Water System i
DESCRIPTION:
The purpose of this Special Test Procedure was to conduct a trial use of a chemical dispersant -(non-ionic polymer antifoulant) to clean the turbine generator condenser tubes by intermittently adding this chemical to the Circulating Water
-(CW) System downstream of the traveling screens, and determine its effectiveness.
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A SAFETY.
ANALYSIS:
The-Circulating-Water System is not safety related and is not credited with a safety function in any USAR accident or transient analysis.
The only effect of this STP on plant operation was to potentially increase 'the efficiency of the turbine generator condenser by reducing fouling of the water side of the condenser tubes. The addition of the antifoulant will compensate for the reduced scouring action from river sand during the winter months.
This STP did not effect station safety, operation or the function of any safety related equipment, and did not involve an unreviewed safety question.
STP 88-194 TITLE:
PMIS Digital Point Overcurrent DESCRIPTION:
The purpose of this Special Test Procedure was to monitor the 28 volt power supplies that supply sensing voltage to the digital input points of the Plant Management Information System (PMIS) Multiplexer (MUX) Links. This provided data, to allow for troubleshooting and repair of intermittent short circuits that have been causing the power supply fuses to open.
SAFETY l
ANALYSIS:
This Special Test Procedure was intended only to monitor the 28 volt power supplies to the PMIS-MUX-LINKS to provide data for troubleshooting.
The affected PMIS multiplexers do not perform a
safety function nor impact any Technical Specification. The PMIS computer system is part of the plant monitoring system only, and does not dir3ctly affect the performanae or operation of any plant system including those important to safety. Therefore, performance of this STP did not involve an unreviewed safety question.
STP 88-201B TITLE:
Monitoring of Steam Tunnel Cooling System Temperatures
-DESCRIPTION:
The purpose of this Special Test Procedure was to obtain ambient air, cooling water, and surface temperature associated with the cooling system and heat load contributors in the Steam Tunnel. Thermocouples were temporarily installed in the Steam Tunnel to achieve this purpose.
The information obtained from this STP will be used in evaluating Steam Tunnel cooling system performance.
SAFETY ANALYSIS:
The Steam Tunnel Cooling System which is part of the Turbine Equipment Cooling (TEC) System is a non-essential, non-safety related system.
Thus, this STP did not affect any systems performing t.
safety lunction.
The function of the Steam Tunnel cooling system is to provide normal operation cooling of the area. The only purpose of this STP was to monitor and obtain data for system performance evaluation. There was no changes in system components or system operating characteristics, therefore, the af fect on overall plant safety was not changed.
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STP 88-286
-TITLE:
Radiation Monitor Flow Switch (RMP-FS-332) Replacement DESCRIPTION:
The purpose of this Special Test Procedure was to provide for the temporary installation and performance testing of a replacement component for RMP FS-332, Service Water effluent radiation monitor sample flow switch.
SAFETY ANALYSIS:
The flow switch involved with this STP was non-essential, and is not required to perform any safety-related function. The replacement flow switch for the Service Water effluent is anticipated to improve overall monitor. reliability.
Performance of this STP was in conformance with plant Technical Specifications which require that process radiation monitoring equipment shall be operable; except for maintenance, and required tests, checks, and calibration.
l Therefore, this STP did not involve an unreviewed safety l-question.
STP 89-091 TITLE:
Pall Condensate Filter Septum Performance Evaluation DESCRIPTION:
Special Test Procedure 89-091 provided for the installation j
l and performance evaluation of Pall Corporation's " Profile" filter septa which was installed in the "C" and "D" condensate demineralizer vessels. This STP evaluated the septa for resin
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leakage, soluble and insoluble metal removal, pressure drop before and after precoat, total volume of water processed between precoats, and septa life span.
SAFETY ANALYSIS:
The "C"
and "D"
condensate filter septum replacement was necessary due to partial filter pluggage and was considered routine maintenance. The Pall septa were a direct retrofit of the existing Graver demineralizer.
Therefore, no vessel modification was required prior to installation / replacement.
The Filter septa were removed and replaced utilizing CNS
- Procedure 2.2.5 " Condensate Filter Demineralizer System", and filter performance data was collected utilizing CNS operating and chemis try procedures. Therefore,-all margins of safety as defined by Technical Specification. USAR, and plant procedures
'were maintained.
This STP did not involve an unreviewed safety question.
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STP 89-212 i
TITLE:
Measurement of Air Flow & Air Temperature of the Control Room DESCRIPTION:
The purpose of this Special Test Procedure was to measure and record air flows and temperature data in the Control Room and Cable Spreading Room under normal operating conditions. This data will than be used to calculate the Control Room Envelope internal heat load. This calculation will be compared to the original heat load calculation to determine if additional heat load has been added to the Control Room Envelope since initial plant start up.
SAFETY ANALYSIS:
This STP took air flaw and temperature measurements in the Control Room and Cable Spreading Room areas.
The STP used a pitot tube, manometer, and thermometer to measure HVAC System air flows and temperature.
The measurements were taken at selected locations in the Control / Cable Spreading Room (Control Room Envelope) during normal operating conditions In addition, temperature in the rooms was monitored at 60 minute intervals during performance of the STP.
Thorofore, operability of all essential and non-essential components were not af fected by the performance of this STP. This STP did not increase the probability or consequence of any accidents or possibility malfunctions previously evaluated, nor create a for an accident or malfunction of a different type than previously evaluated in the USAR.
All applicable plant Technical Specifications pertaining to the Control / Cable Spreading Rooms and associated essential equipment, and room air flow /tenperature were maintained at all times. Therefore, all margins of safety as defined in the Technical Specifications were maintained.
This STP did not involve an unreviewed safety question.
STP 89-246 TITLE :
Primary Cont. Purge & Vent Valves Quarterly Leak. Rate Tests DESCRIPTION:
The Purpose of this Special Test Procedure was to provide guidance for performing quarterly leak rate tests on the primary containment purge and vent isolation valves.
The valves are normally leak tested once per operating cycle per 10CFR50 Appendix J.
However, NRC concerns have prompted increased frequency of the tests to verify the adequacy of the valves' resilient seats.
SAFETY ANALYSIS:
Since the tests described in this Special Test Procedure were performed with both the inboard and outboard valves of each purge and vent penetration in the closed position, the quarterly leak tests did not degrade the Primary Containment isolation capability of the subject valves. This STP did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration.
Therefore, the af fect on overall plant safety was not changed.
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'STP 90-174J TITLE:
Loss of Gland Water to the Service Water. Booster Pump Functional Test DESCRIPTION:
The~ Purpose of this Special Test Procedure was to determine what effect loss of gland cooling water would have on the Residual lleat Removal (RllR) Service Water Booster Pump (SWBP) mechanict' seals, SAFETY ANALYSIS:
This STP was performed just prior to a scheduled maintenance inspection on a selected SWBP. The maintenance inspection and the performance of this STP required entering a Limiting Condition of Operation (LCO) while.the.SWBP was out 'of service. This STP did not adversely impact any of the design features of the Service Water System and the STP provided for sufficient monitoring of the pump to detect and prevent any damage..This STP was performed in compliance with the LCO as specified in the CNS Technical Specifications. Therefore, by complying-with Technical Specification's operability requirements, the probability of occurrence or. the consequences of an accident or malfunction previously evaluated was unchanged. The margin of safety as used in the basis for any Technical. Specification was -not reduced, therefore, an unreviewed safety question did not exist.
STP 90-269 Amendment 1 TITLE:
Primary Cont, Purge & Vent Valves Quarterly Leak Rate Tests
-DESCRIPTION:
The Purpose of -. this Special Test Procedure was to provide guidance for performing quarterly leak rate tests -on the l
primary containment - purge and vent isolation valves.
.The valves are normally leak tested once per operating cycle per
-10CFR50 Appendix J.
However,- NRC - concerns have prompted increased frequency of the tests to verify the adequacy of the resilient seats. In addition, STP 90-269 Amendment I replaced STP 90-269 " Primary Containment Purge and Vent Valve Quarterly Leak Rate Tests" in its entirety.
ls SAFETY ANALYSIS:
Since the tests describad in this Special Test Procedure were performed with both the inboard and outboard valves of each.
purge. and ~ vnnt penetration in the closed position, the quarterly-leak tests did not degrade the Primary Containment isolation capability of the subject valves. This STP did not require abnormal operation of any plant systems or pr.,cedures, and did not introduce any plant equipment alteration.
Therefore, the affect on overall plant safety was not changed.
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REPORTABLE DESIGN CHANGES-DC's 87-015MA. MB. MD. and ME TITLE:
CNS Annunciator Upgrade Project DESCRIPTION:
These Design Changes were a continuation of the Detailed Control Room Design Review (DCRDR) Annunciation Upgrade Proj ec t.
This Project replaced the existing non-essential (not safety related) Cooper Nuclear Station (CNS) Control Room Annunciator System with an upgraded non-essential Annunciator System which incorporates reflash and sequential events recording capabilities and is designed to meet Human Factors Engineering (HFE) guidelines.
These Desig.,
Changes accomplished the replacement of the existing Panalarm Control Room yindowboxes-in the following Control Room Panels (A, E, H, K, M, P-1, P-2, Q, R and 9-3) with new windowboxes, Control Room Supervisor's CRT, Alarm printers, Panel CRT's, and the necessary connections.
SAFETY ANALYSIS:
These Design Changes improved the annunciator system performance and reliability as well as resolved Human Factors Engineering deficiencies. These Design Changes did not affect the actual systems (alarm inputs) that the affected Panels monitor. The performance and reliability of the systems which provide input to the Annunciator system were not changed by these Design Changes, and the Plant was in a cold shutdown condition while the work was implemented.
The Annunciator System is a non essential system and its operaticn, although desirable, is not necessary to obtain safe shutdown of the plant.
A failure of any portion, or the entire Annunciator System will not jeopardize the plant safe shutdown capabiliti.
Although the Annunciator -System interfaces with safety systems, neither an Annunciator System failure or an interface failure will jeopardize the plant safety system capabilities.
These Design Changes did not modify the function of any safety system. The upgraded Annunciator System only utilizes passive monitoring of existing instrumentation contacts. In addition,-
all applicable Technical Specifications were adhered to during litiplementation of these Design Changes.
Therefore, these modifications did not change the existing accident analyzes for Cooper Nuclear
- Station, nor the probability or _-
-consequences of an accident as analyzed in the USAR.
No reduction in the margin of safety was involved with implementation of these Design Changes.
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DC 87 023 TITLE:
PAD Chemical Laboratory Modificaticus DESCRIPTION:
This, Design Change performed environmental and instrumentation modifiestions to the Radiological Chemical Laboratory in the Radwaste Building at Cooper Nuclear Station.
In addition, this Design Change provided for irnproved monitoring for the Condensate Filter Demineralizer Cot.ductivity Sampling System as well as improved environn. ental and space conditions for the RAD Chemical Laboratory personnel.
SAFETY ANALYSIS:
This Design Change did not affect any safety related systems.
nor did v affect the safe optration or shutdown of any essential system, and was classified as non essential. Since none of the changes associated with Design Change 87 023 affected any safe shutdown systems or components and the quality of materials were equal to or greater t hr. n those specified in the original construction, implementation of this design change did not increase the probability of occurrence or the consequence of an accide. t or malfunction of equipnet 1
trnportant to safety previous's eunluated in the USAR.
In addition, this desibn ch. y did tot require any changes or additions to the Technical Specifications and involved no decrease in the margin of safety.
Therefore, no unreviewed
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safety question existed with the implementation of this design change.
DC 87-023 Amendment 1 TITLE:
RAD Chemical Laboratory Modifications - Amendment 1 DESCRIPTION:
The purpose of this Amendment to Design Change 87-023 was to provide for the furthur expansion of the Radiological Cnemical Laboratory Office located in the Radwaste Building at Cooper Nuclear Station.
SAFETY ANALYSIS:
This Amendment did not invoh 3 modifications to safety equipment and was. therefore, i_onsidere4 non essential.
The implementation of this araendment dio not degrade plant personnel safety, equipment safety or nuclear safety during or following the modifications. This Amendment did not increase the possibility of any accident occurrence, nor did it decrease the safety margin as defined by the basis for any Technien1 Specification, therefore, it did not involve an unreviewed safety question.
DC R8-053B TITLE:
Essential Control Building Ventilation System DESCRIPTION:
This Design Change (DC) installed an essential (safety related) ventilation system to the Critical Switchgear Rooms and to the Control Building (903'6" elevation) to provide 13
I cooling under abnormal and accident conditions. This change utilized a network ot essential supply and exhaust fans to remove the heat generated by both the essential electrical equipment and normal equipment in the control building. This system was clar,1fied as a ventilation system because no chilled water or mechanical refrigeration were involved. This DC fulfilled the District's commitment made in response to NRC Inspection Report 50 298/89 13, dated April 10, 1989, to install the essential ventilation system for cooling the Critical AC and DC electrical equipment rooms under abnormal and accident conditions, i
SAFETY r
ANALYSIS:
The new Essential Ventilation System was designed for l
redundancy, single failure criteria, separation criteria, and other rules applicable for a nuclear safety related system, This new ventilation system will provide a safety function in that it wiD maintain acceptable room t a peratures to further ensure that the necessary electrical equipment is available to safely shutdown the reactor, No new safety concerns were created with the implementation of the Design Change, in addition, CNS plant safety was enhanced by the installation of the Essentini Control Building Ventilation System. 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 Design Change.
DC 88 201 B T?TLE:
Steam Tunnel Cooling System Upgrade l
DESCRIPTION:
This Design Change modified the Steam Tunnel Cooling System.
The modifications included the installation of replacement fan coil units (FCU), and the addition of insulat:on to the Main Steam Line penetration gusrd sleeves in the iteam Tunnel to reduce the Steam Tunnel heat load.
The Stem Tunnel FCU's re'lacement and insulation installation were performed to dt m easo area temperature during normal operation.
SAFETY ANALYSIS:
The implementation of this Design Change did not de;;rado plant personnel safety, equipment safety, or nuclear safety.
The Steam Tunnel Cooling s'/ s t em is non essential and is not required for safe shutdown of the plant, or mitigating the consequences of an accident. The ability of the Steam Tunnel I
Cooling system to perform its function was increased by this modification.
The performance of the cooling system was enhanced by an _ increase in system heat load removal capabilities and by the reduction ir. heat input from the Main Steam Line penetration guard sleeves. The design function and operation of the Steam Tunnel Coolirg system remained unchanged.
Therefore, this Design Change did not create an unreviewed safety question or have an adverse effect on nuclear safety.
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4 DC 89 002 TITLE:
West Varehouse Utilities Connection DESCRIPTION:
The purpose of this Design Change was to connect the new West Warehouse at CNS to the existing fire potable !!02 protection 2
lines, Caitronics intercommunications system, and the 12.5kV ring bus power line.
In addition it also provided for the installation of the sprinkler system in the building and the l
fire alarm annunciator panel in the Central Alarm Station.
SAFETY ANALYSIS:
The roodifications outlined by this Design Change did not degrade the safety of Cooper Nuclear Station with respect to equipment or nuclear safety. This change did not change the function or operation of any system or component related to safo shutdown of the plant.
This DC did not create an unreviewed safety question, nor did it reduce the margin of safety defined in the Technical Specifications.
t DC 89-180 TITLE:
Testable Check Valve Actuators DESCRIPTION:
The purpose of this Design Change was to improve the testing which ensures the operability of the Core Spray (CS) and Residual lleat Removal (pdlR) systems testable check valves.
Thi. was accomplished by removing the air operated actuators and the motor operated bypass valves (CS system only) associated with these testable c),eck valves. Modifications to the RilR and CS testabic check valves are consistent with the NRC recommendations per AE0D/C502, "0verpressurization of Emergency Core Cooling Systems for Boiling Water Reactors".
l The motor operated bypass valves (CS system only) were no longer required once the air operators were
- removed, therefore, they were also removed.
In addition, the disc position indicators were modified to improve the valvo position indication reliability.
SAFETY ANALYSIS:
The implementation of this DC was performed while the plant i
was in a cold shutdown condition. This DC did not change the original design basis of the testable check valves or affect the safety function of the affected systems.
However the changes did improve the reliability of the CS and RHR testable l
check valves by modifying the reed switches for more reliable position indication. In addition, the CS system bypass valves
_ piping and instrument ait drywell penetrations wece cut, capped, and hydrostatically (leak) tested. This Design Change did not alter the capabilities of the CS ot RHR testable check i
valves, nor did it change any functions of the affected components during operation.
The margin of safety was not reduced nor was the possibility of an accident or malfunction created or increased by the implementation of this Design Change.
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DE 89-207 TITLE:
!!a g Analyzer particle Filters Isolation Valve 14odifications 0
DESCRIPTION:
The purpose of this Design Caange was to install four particle filters in the sample lines to the H 0, Analyzer Units to 2
improve Analyzer operation. Four additional isolation valves were installed in the sample lines to provide Operations a simpler way to isolate the drywell and torus from the H 0 2 2 Analyzer - Units.
Two shutoff valves were installed in the return lines from the 110 analyzers to the torus to provide 2 2 a simpler way for Operations to isolate the Analyzer " nits.
Finally this DC installed four sample point valves in the sample lines to provide a casier way to perform grab samples when the 110 Analyzer 1mits are inoperable.
2 SAFETY ANALYSIS:
The Containment Atmosphere Monitoring System operation did not change with the implementation of this Design Change.
The installation of the particle filters provide additional assurance that abrasive particles will not enter the analyzer units.
The implementation of this DC increased system reliability, efficiency, and operator performance.
Installation of the filters and valves did not affect the 110, 2
monitors, in that the system will retain its safety features and that operation of the system during a design basis event remains as specified.
This DC was implemented with the Reactor in a cold shutdown condition, This DC did not constitute an unreviewed safety question, nor did it reduce the margin of safety defined in the Technical Specifications.
DC 89 256 TITLE:
Reactor Water Cleanup (RWCU) Pipe Replacement DESCRIPTION:
The purpose of this Design Change (DC) was to replace portions of the Reactor Water Cleanup (RWCU) piping system with material that is resistant to intergranular stresa corrosion cracking (ICSCC).
In addition, this DC implemented several minor modifications (upgrade of the RWCU pumps, addition of a subcooling line, bypass lines, and replacement of excess flow elbow taps with an Annubar flow element) to improve system performance and reliability.
SAFETY ANALYSIS:
Thfs Design Change enhanced the existing plant design by upgrading RWCU pump performance, replacing portions of the RWCU piping and adding bypass lines, subcooling lines, annubar flow elements etc..
These changes resulted in several benefits including sin.plification of operation, reduced maintenance, and reduced radiation exposure.
These changes were implemented while the plant was in a cold shutdown condition. No safety dasign basis or functional requirements of the systems were affected.
Therefore, this modification did not change the existing safety analysis for Cooper Nuclear Station, nor the probability or consequence of an accident as analyzed in the CNS USAR.
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DC 89-272 TITLE:
Combustible Gas Control - Standby Nitrogen Injection (SBNI)
Standby Nitrogen Inj ection DESCRIPTION:
This Design Change provided a (SBNI) System for injecting nitrogen Sas into the Primary Contaitunent.
The purpose of this nitrogen injection is to provide an emergency backup sys tern to the normal Primary containment nitrogen supply system.
These systems are to dilute the combustible gases produced by radiolytic l
decomposition of reactor coolant and metal water chemical reactions which occur following a Loss of Coolant Accident (LOCA). The SBNI system will provide the necessary supply of 7
nitrogen for a minimum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> by which time an off site long-term source of nitrogen will be available.
The conceptual design, system classification, and other details were given in a Letter from NPPD to NRC dated September 28, 1989, " Post Accident Combustible Gas Control".
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SAFETY ANALYSIS:
The SBNI system is provided with two independent connection points, and two independent and redundant injection paths into both the drywell and wetwell (suppression chamber) portions of the CNS Primary Containment.
The existing purge and vent system was the prirnary means of nitrogen inerting. and was therefore subjected to gaseous nitrogen under normal conditions. The replacement of the existing ACAD system with the new SBNI system will enhance the ability to inj ect dilution gas into the primary containment during accident conditions. The SBNI system is provided with two complete and i
redundant pathways and nitrogen sources ensuring against a j
single component or pathway failure which would prevent the overall system from performing its function. This DC did not t
i constitute an unreviewed safety question, nor did it reduce the margin of safety defined in the Technical Specifications.
I DC 89 286A TITLE:
Performance / Reliability Monitoring Instrumentation i
DESCRIPTION:
This Design Change installed instruments to monitor the performance and reliability of safety related components. The safety related components included the Diesel Generators, the Residual Heat Removal System, and the Diesel Generator Service Water System.
These instruments were included in the Performance / Reliability Monitoring Program implemented at CNS.
The installation of those instruments will allow for the collection of data required for in depth analysis of system components and their performance. This analysis will further ensure the reliability of the safety systems by providing valuable trending data.
SAFFTY ANALYSIS:
No personnel, equipment, or nuclear safety concerns existed with the implementation of this DC. The components installed per this DC are passive components (local indicators only) and do not provide any controlling functions.
The addition nf 17
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these instruments-does not atfect the operational characteristics of any of the safety related or non essential components / systems.
Since. there was no change in any system components or operating characteristics, the effect on overall plant safety was not changed.
DC 89 289 TITLE:
Instrument Air (IA) Post Filter "B" Replacement DESCRIPTION:
The purpose of this Design Change involved the replacement of the Instrument Air Dryer B Post Filter, and the replacement of the existing tilter cartridges on both Trains A and B with high temperature cartridges.
In addition, piping was rerouted, a filter blowdown valve was added and a connection for a temperature indicator was installed.
SAFETY l
ANALYSIS:
This Design Change did not create the possibility of an -
accident or malfunction of a different typo than previously identified. The portions of the affected system that provide engineered safeguards and reactor protection functions were not altered by this DC.
The Instrument Air system design, operation, capacity and, capability were not degraded by this DC. The Instrument Air system operates in the same manner as it did prior to the modifications. Therefore, this DC did not constitute an unreviewed safety question, nor reduce the margin of safety as defined in the Technical Specifications.
DC 90 00/1 TITLE:
Non Critical AC Bus Coordination DESCRIPTION:
Design Change DC 90-004 involved modifying - various non-i critical 4160V relays and 480V circuit breakers by changing the setpoints,_ refurbishing solid-state trip devices and changing their setpoints,;or replacing the circuit breakers
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with fused disconnects.
The changes were based on the analysis of the Non-Critical AC Buses at Cooper Nuclear Station by Nebraska Public' Power District.
These changes provide improved breaker coordination = and will reduce the possibility of falso trips during fault conditions.
SAFETY ANALYSIS:
Equipment reliability was enhanced by.the implementation of this Design Change. The modifications performed provided_ for adequate circuit coordination margins on non-critical AC buses which supply power to non-safety related loads at CNS.
With the new coordination margins, a failure of any equipment, component, or cable will isolate the failure to a minimum area, thereby, decreasing the probability that the failure of the equipment or cabic will affect additioral equipment. All work done by this DC was done in accordance with established plant procedures for setting and/or testing braakers.
No unreviewed safety questions were created, nor was any Technical Specification margin of safety reduced.
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DC 90 021 TITLE:
PMfS Augmentation, Phase II DESCRIPTION:
The purpose of this Design Changa was to provide additional plant parameter inputs to the Plant Management Information j
System (PMIS), to correct human factors deficiencies observed l
during the Detailed Control Room Design Review (DCRDR), and to i
comply with truREG 1342 requirements for irtplementation of Safety Parameter Display System (SPDS). In addition, this DC imphmented additional FMIS points to improve monitoring of system operation.
SAFETY ANALYSIS:
Although the PMIS interfaces with safety systems, PMIS failure will not jeopardize the plant safety system capabilities. All interfaces with safety systems utilize essential components.
This Design Change did not modify the function of any safety syt. Le m. The upgraded PMIS only utilizes passive monitoring of-existing instrumentation contacts.
Therefore, this modification did not change the existing accident analyses for Cooper Nuclear Station, nor the probability or consequences of an accident as analyzed in the USAR.
No reduction in the margin of safety was involved with implementation of this l
Design Change.
l DC 90 181 TITLE:
Modification Of DC Westinghouse DB Breakers DESCRIPTION:
The purpose of this Design Change was to convert the existing Westinghouse DB series circuit breakers into fused disconnect switches.
The conversion and sizing of the fuses in the modified breakers were changed for DC breaker coordination purposes.
In addition, fuse status indicating lights were installed (with ann.inciation provided) in the door of the DC l
Switchgear cubicles to indicate blown fuses, i
SAFETY ANALYSIS:
The work involved in this Design Change did not result in any personnel, equipment, or nuclear safety problems, nor did it result in any operational changes for the systems affected.
The use of fuses for electrical coordination in DC electrical systems is superior to circuit breakers for protection against switchgear faults therefore, system reliability was increased.
The work for this DC was performed in the plant's Electrical-shop, where the breakers were modified and acceptance tested prior to reinstallation. The safety function of the DC system was enhanced and the ability of the DC batteries to power safety-related components during a Design Basis Event was not l
changed. Therefore, the affect on overall plant safety was not changed.
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DC 90-218-TITLE:
Residual lleat Removal (RllR) Pressure Maintenance Test cauge i
DESCRIPTION:
The purpose of the this Design Change was to provide a positive means for pressure raonitoring of Check Valves allR CV-24CV and RllR CV 25CV in the RilR "A" Loop and Check Valves RHR-CV 18CV and RllR CV+19CV in the RilR "B" Loop as required for ASME Section XI In Service Testing. This was accomplished by installing tubing and a pressure gauge to allow for testing of the valves.
SAFETY ANALYSIS:
The affected portion of the system where the modifications were involved was the non essential portions of the system.
This DC installed tubing and a now pressure gauge that is isolated by a normally closed root valve.
These components are in service only during periods of Inservice Testing and are isolated during normal plant operation. The installation
- of the pressure gauge provides for positive monitoring of the check valve performance during the required testing.
The overall reliability and the safety design function of the check valves has not been altered. The-implementation of this DC did not affect the probability 4 occurrence or the consequences of an accident or malfunction of equipment.
important to safa y, nor did it create an accident or malfunction of a different type as defined by the USAR and Technical Specifications.
DC 90 302 TITLE:
ASCO Transfer Switch Modification DESCRIPTION:
This design Change was required to close out Temporary Design Change (TDC) 90 300 which allowed Motor Control-Center (MCC)
MCC T to be the normal feeder for MCC X with MCC L as the emergency feeder thereby, providing a more appropriately distributed load on the diesel generators.. In addition, this l
DC also removed the unused automatic transfer logic on some ASCO model 935 307 transfer switches.
The recoval of the unused circuitry will allow for more reliable operation since the new design will contain fewer components with less maintenance.
SAFETY ANALYSIS:
Tuis DC reduced the ECCS loading on DG #1, and increased the ECCS loading on DC #2, which was not as heavily loaded during postulated _ accident scenarios.
This DC ensured an adequate source of electrical power to operate the essential equipment l_
by reducing the peak loading of DC ul. The additional loading on DG #2 was well within its rated output.
The manuni operation of the ASCO transfer switches did not change, only the unused components of the automatic transfer switches were removed.
This DC did not affect tbo normal operational or maintenance characteristics of the ASCO transfer switches.
Therefore, no margin of safety was increased, and no unroviewed safety question existed.
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ESC 91 045 TITLE:
Non Critical AC Fuse Replacement DESCRIPTION:
The purpose of this Equipment Specification Change was to replace the existing fuses which did not conform with the Non.
Critical AC 15us Coordination Study, Nucicar Engineering Department Calculation (NEDC) NEDC 86-105F. These fuses were changed out to match the coordination study and provide standardization on fuse brand and types used at CNS.
SAFETY ANALYSIS:
The new fuses that were installed have the same physical dimensions, are capabic of handling the full load and inrush currents of the
- circuits, will provide short circuit protection, and will provide the proper fuse coordination.
With the new coordination margins, a failure of any piece of equipment or cab'e will isolate the failure to a minimum area, thereby, decreasing the probability that the failure of the equipment or cable will affect additional equipment.
Therefore, since the design and operation of the systems involved were not changed by this ESC, no unreviewed safety questions were created.
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REPORTABLE ACTIVITIES Setooint Chanre Reauest 89-26 TITLE:
ARI/RPT ATWs H16h Reactor Pressure Trip Setpoint Change DESCRIPTION:
The purpose of this setpoint change to the high pressure Recirculation Pump Trip (RPT) and Alternato Rod Inrertion (ARI) was to ensure that ARI/RPT will occur at Cooper Nuclear Station to provide Anticipated Transient Without Scram (ATWS) protection when initiated at lower power levels, but will mini'nize the potential for inadvertent occurrence.
This j
setpoint change (89 25) raised the setpoint of Pressure Switches NBI PS.102A, B, C and D from 1060 psig to 1074 psig.
The selection of the reactor high pressure setroint (1074) for i
i the ARI/RPT logic of the Cooper Nuclear Station is described in General Electric Company Document EAS+16 0389, " Evaluation of ARI/RPT High Pressure Setpoint for Cooper Nuclear Station".
SAFETY ANALYSIS:
This activity involved only a setpoint change and did not involve any modifications to the Reactor Recirculation System or its control system.
Therefore, this activity did not change the safety function of the Reactor Recirculation System or its performance and reliability. The modification did not involve any changes to ARI/RPT system hardware.
The CE Analysis, EAS 16-0389, utilized normal input parameters 75'F) rather than (e.g.,
service water temperature conservative-bounding conditions.
ATWS involves multiple failures, therefore, more realistic evaluations are acceptable than in the case of design basis accidents discussed in the USAR, which analyzes single failure with subsequent mitigating backup and redundant components.
Accordingly, it was acceptable to hdC that the analysis input parameters be less conservative than USAR values.
The new setpoint for high pressure AR1/RPT is below the Technical Specifications limit of 1120 psig for Cooper Nuclear Station RPT.
This'setpoint change did not therefore, present an unreviewed safety question and did not require a change to the CNS Technical Specifications.
Proacdure Chance Notice (PCN) 0.3 (Revision 11)
TITLE:
Station Operations Review Committee _
DESCRIPTION:
The purpose of this Procedure Change Notice was to document the addition of the Senior Manager of Staff Support position to the CNS Station Operations Review Committee (SORC) membership. This change provided for additional expertise in the diverse areas associated with nuclear plant operation. In addition, this change increased the SORC Quorum requirements from five to six.
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SAFETY ANALYSIS:
This procedare change did not in any way degr9de the safety of Cooper Nuclear Station with respect to personnel, equipment, l
or nuclear safety, The change was administrative in nature i
i and involved no technical or operational aspects that directly affect station operation.
This procedure change did not require abnormal operation of any plant systems or procedures, i
and did not introduce any plant equipment alteration.
Therefore, the effect on overall plant safety was not changed.
i Procedure Chanr_e Notice (PCN) 2.2.3 (Revision 40)
TITLE:
Circulating Water System DESCRIPTION:
The purpose of this procedure change notice was to add l
administrative liini t s in the Circulating Water Sys tern procedure restricting the backwashing of both condensers while the mode switch is in RUN.
During power operations.
l backwashing both condensers at the same time under certain conditions could result in a significant loss of condenser vacu tm and a possible plant shutdown.
SAFETY ANALYSIS:
The Circulating Water System is not relied upon to mitigate the affects of any analyzed transients or accic Les, and system failure would not result in an unanalyzed accident.
However, by adding administrative limits restricting certain backwashing activities, plant safety was enhanced by removing a potentini plant transient, resulting from a loss of condenser vacuum. There was no change in system components or system operating characteristics, therefore, no unreviewed safety questions were created.
Procedure Chance Notice (PCN) 5.2.5 (Revision 20)
TITLE:
Loss of Normal AC Power - Use of Emergency AC Power DESCRIPTION:
This procedure change notice was a direct result from actions taken by NPPD in response to NRC Information Notice (IN) 86-70
" Potential Failure of All Emergency Dicael Generators". This IN discussed design deficiencies that could disable both diesel generators (DG) by placing unanalyzed loads on the DG powered buses.
l The CNS review used GE Specification 22A1259, " Standby AC Power System" to determine the ECCS equipment operational requirements in - addition to the electrical drawings to determine which loads automatically shed during transfer to the emergency transformer or diesel gere: ators (DG's).
This PCN accentur sd the loading restrictions on the. DG 's, to prevent overloading during the initial two (2) hours when the maximum ECCS load is present, and to clarify the load shed details of the BUS 1F and 1G transfer from normal power to the emergency transformer. All essential loads were analyzed for j
adequate voltage and current when the non-essential MCC's are powered from the emergency transformer for both high and inw line voltage and load conditions.
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l SAFETY ANALYSIS:
This PCN ensured that the 4160V 4.nd 480V auxiliary power distribution system will, under all transient and accident conditions, distribute AC power reauired to safely shutdown the reactor, maintain the safe shutdo in condition, and operate all auxiliaries necessary for station safety by placing loading restrictions on certain non essential loads. This PCN also provided administrative guidance to ensure that the DG's are not overloaded during the initial two (2) hours in an accident scenario when the maximum ECCS load occurs, and clarifies the load shed details of the BUS 1F and 1G transfer from normal power to the emergency transformer. In addition, this PCN ensures that an adequate source of electrical power t c. operate the essential equipment is available by placing loading restrictions on certain non-essential loads which increases the margin of safety for the diesel generator's loading capability.
Therefore, the diesel generator's reliability, and capability to perform its intended safety function was not jeopardized.
Other Activlt h TITLE' Evaluation of Dropped Control Blade in Spent Fuel Storage Pool DESCRIPTION:
This activity involved the movement of Control Rod Blades into and out of the Spent Fuel Storage Pool during a refueling outage along restricted paths in the vicinity of irradiated fuel without the benefit of Secondary Containment.
An analysis was performed with respect to resulting fuel damage in the event a Control Rod Blade was dropped near irradiated fuel and subsequently impacted the fuel to determine if Secondary Containment was required for this activity.
SAFETY ANALYSIS:
The analysis (GE proposal 295-1CB6L-HP1 91) showed that the energy from a falling control blade was insufficient to fail any fuel rods in the struck bundle when the blades were restricted to certain areas. Consequently, there would be no release of radioactivity from fuel bundles hit by a control blade that dropped vertically to the fuel rack and then fell over onto the stored fuel.
Based on the results of this evaluation (and controlled movement of the refueling machine),
it was concluded that if the arrangement of irradiated fuel bundles in the spent fuel storage pool are positioned so that a control blade car.not be directly dropped on the fuel during the blade movement through the pool, then this specific refueling operation would not require Secondary Containment integrity to be maintained during this activity.
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PERSON.
AND MAN-REM EXPOSURE I
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25
PER$0N?EL AND MAN. REM BT WORK AND JOB FUNCTION i
itunber of Parsom3L
(> 100 mrem)
Total Man tem Station Utility Contractor &
Station Utility Contractor &
W;rk and Job function Emloyees Emloyees Others Emleyees E mlovees others REACTOR OPERATIONS & SUPV.
taintenance Personnel 7
0 0
0.337 0.000 0.000 Operating Persomel 53 0
0 18.272 0.000 0.000 r:stth Physics Persomel 25 0
37 8.284 0.000 11.753 Supervisory Personnel 4
0 0
0.913 0.000 0.000 Engineering Personnel 27 1
11 6.238 0.060 1.893 Pot / TINE MAfWtf M NCE Melntinance Persomet 90 0
335 66.264 0.000 184.946 Operating Persomel 0
0 0
0.000 0.000 0.000 M:31th Physics Personnel 31 0
37 21.135 0.000 11.693
)
Supervisory Personnet 0
1 1
0.000 0.340 0.203 Engimer8ng Persomet 8
33 33 0.216 18.177 8.065 SPECIAL MAINTENANCE Kalntsnance Persomet 1
0 20 0.018 0.000 13.463 operating Persomet 6
0 0
0.039 0.000 0.000 Health Physics Persomel 2
0 8
0.184 0.000 1.592 supervisory Persomel 0
0 0
0.000 0.000 0.000 Engineering Persomel 3
0 1
0.124 0.000 0.170 WA$TE Pe0CTfSING Kaintenance Persomet 1
0 0
0.004 0.000 0.000 Operating Personnel 4
0 0
1.625 0.000 0.000 H1alth Physics Persomet 5
0 1
1.103 0.000 0.017 Supervisory Persomet 0
0 0
0.000 0.000 0.000 Engimering Personnel 0
0 0
0.000 0.000 0.000 REFUELING Malntenance Persomet 0
0 0
0.000 0.000 0.000 Operating Personnel 31 0
0 1.010 0.000 0.000 Histth Physics Personnel 2
0 2
0.233 0.000 0.237
$upervlsory Personnel 0
0 0
0.000 0.000 0.000 EnginJering Personnet 2
0 0
0.089 0.000 0.000 INSERVICE INSPECTION Xalntenance Persomel 1
0 21 0.002 0.000 11.755 Operating Persomel 0
0 0
0.000 0.000 0.000 Ntalth Physics Personnel 0
0 2
0.000 0.000 0.095 Supervisory Ph 'onnel 0
0 0
0.000 0.000 0.000 Enginstring Persomet 1
0 0
0.001 0.000 0.000 191Ak Maintenance Personnel 90 0
375 66.625 0.000 210.164 l
Operating Persomet 55 0
0 20.946 0.000 0.000 Msslth Physics Persomel 33 0
39 30.939 0.000 25.387 Supervisory Persomet 4
1 1
0.913 0.340 0.203 Engineering Personnel 27 33 40 6.668 18.237 10.128 GRAND TOT.'.t$
209 34 455 126.091 18.577 245.882 l
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