ML20012G563
| ML20012G563 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1992 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD930022, NUDOCS 9303090235 | |
| Download: ML20012G563 (19) | |
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i GENERAL OFFICE P.O. BOX 499. COLUMBUS, NEBRASKA. 68602-0499 Nebraska Public Power District
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l NSD930022 March 1, 1993 l
U.S. Nuclear Regulatory Commission 6
Document Control Desk l
Washington, DC 20555 l
l Gentlemen:
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Subject:
Annual Operating Report Cooper Nuclear Station i
NRC Docket No. 50-298, DPR-46 i
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, 1992, through December 31, 1992.
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.
Si cere y, l
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G Horn Nuclear Power l
Group Manager GRH/tja:92an-rpt.ltr Attachment cc:
NRC Regional Office l
Region IV NRC Resident Inspector Cooper Nuclear Station REIRS Project Manager (w/ Personnel Man-Rem Report only)
Office of NRR - USNRC Washington, DC l
l 9303090235 921231 PDR ADOCK 05000298 I
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ANNUAL OPERATING REPORT JANUARY 1,1992 THROUGH DECEMBER 31,1992 USNRC DOCKET 50-298
TABLE OF CONTENTS SECTION PAGE I.
PERFORMANCE CHARACTERISTIC....................
1 Fuel Performance..................................
2 MSV and MSRV Failures and Challenges..................
3 11.
FACILITY CHANGES, TESTS, OR EXPERIMENTS REPORTABLE U N D ER 10C FR50.59................................
4 Reportable Special Procedures /Special Test Procedures......
5 Reportable Design Changes...........................
9 Reportable Activities (Setpoint, Procedure Changes)..........
20 111.
PERSONNEL AND MAN-REM EXPOSURE...............
25 By Work and Job Function............................
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PERFORMANCE CHARACTERISTICS i
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FUEL PERFORMANCE Cycle 15 operation continued from January 1,1992, through February 9,1992 Operation of the unit was interrupted on February 10,1992, via a power reduction and manual scram, for the replacement of some 250 volt battery cells. The unit was re-started on February 15,1992, and continued operation until AprH 19, 1992. The unit was shut down on Apri 19,1992, for the replacement of additional 250 volt battery cells.
Startup commenced on April 24,1992, and operation in cycle 15 continued until Septernoer 10,1992. The unit was shut down on September 11,1992, to address a control power design deficiency related to the design basis accident analysis. A Design Change was implemented to resolve the control power design deficiency and the unit was re-started on September 15,1992, and continued operation through December 31,1992.
Cycle 15 off-gas actMty continued at essentially steady state levels with reactor coolant dose equivalent 1-131 equilibrium values and off-gas release rates maintained well within the limits specified by the CNS Technical Specifications. Comparisons of actual control rod densities predicted by computer program calculations at various core exposures indicated no reactMty anomalies of 1% or greater, I
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MSV AND MSRV FAILURES AND CHAlI FNGES (Ref.: NUREG-0737, Action item II.k.3.3) l There were no operational failures or challenges to the Main Safety Valves or Main Steam Safety Relief l
Valves during the operational year of 1992.
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FACILITY CHANGES, TESTS, OR EXPERIMENTS REPORTABLE UNDER 10CFR50.59 i
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l REPORTABLE SPECIAL PROCEDURES / SPECIAL TEST PROCEDURES t
t STP 88-005 and Amendments 1 & 2 l
i TITLE:
Specimen irradiated Capsule Program (Installation / Removal) l DESCRIPTION:
The purpose of this Special Test Procedure (STP) and Amendments was to provide directions i
in the installation and removal of five specimen capsules into the Reactor Pressure Vessel. The intent of this program is to demonstrate improved resistance to corrosion for zirconium alloys, swelling of hafnium, and to irradiation-assisted stress corrosion cracking for special stainless steels.
SAFETY ANALYSIS:
The five irradiation capsules were installed into the existing neutron source holder positions in l
the core. The only impacted system associated with this STP was the reactor core. There is no significant difference between capsule and original neutron source holder and there was no impact on lattice physics calculations. Additionally, analysis performed showed that the specimens would not escape the capsule so no new type of accident was created. 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 it was determined that 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 90-116 TITLE:
Equipment Space Temperature Study DESCRIPTION:
The purpose of this Special Test Procedure was to monitor and record temperature data in areas where the essential and/or Equipment Oualified (EO) steam leak detection temperature
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switches are located. The information obtained from this STP will be used in evaluating steam leak detection temperature switch performance.
SAFETY ANALYSIS:
The temperature data collection performed by this STP did not change or degrade the performance and/or reliability of the Steam Leak Detection system. This STP did not affect any systems performing a safety function. The only purpose of this STP was to monitor and obtain localized area temperature data to aid in evaluating temperature switch performance. There were no changes in system components or system operating characteristics, therefore, the affect on overall plant safety was not changed.
STP 90-167 TITLE:
Testing of lonex Precoat Materials DESCRIPTION:
Special Test Procedure 90-167 provided for the installation and performance evaluation of Toray lonex material that was applied to the *D" Condensate Filter Demineralizer vessel. This STP evaluated the filtration characteristics, ion exchange capacity, and overall run length of the filter demineralizer with the Toray lonex precoat material applied.
SAFETY ANALYSIS:
The *D" condensate filter deminerallzer precoat material replacement was considered routine maintenance. No vessel modification was required prior to installation / replacement of the lonex precoat. The new Torty lonex precoat material was installed utilizing CNS Procedure 2.2.S
- Condensate Filter Demineralizer System" and filter performance data was collected utilizing CNS operating and chemistry procedures. All margins of safety as defined byTechnical Specification, USAR, and plant procedures were maintained.
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STP 90-173 TITLE:
BWROG Supplemental Surveillance Program Capsule Holder DESCRIPTION:
This Special Test Procedure (STP) addressed the installation of a surveillance capsule holder into the Cooper Nuclear Station vessel for the Boiling Water Reactor Owners' Group (BWROG)
Supplemental SurveHlance Program (SSP). The purpose of the holder is to allow the specimens it contains to accumulate neutron fluence for future testing, thus increasing the size of the i
surveillance data base for BWRs.
j SAFETY ANALYSIS:
The SSP holder is designed to withstand all accident, transient, and normal operational events, and as such, has no impact on plant operation or plant safety. The safety analysis evaluated the worst case scenario of failure of the holder causing a loose part. The evaluation demonstrated that even in the worst case scenario, an unreviewed safety issue or a reduction in the safety margin would not occur. This STP did not effect station safety, operation or the i
function of any safety related equipment. The probability or consequences of an accident analyzed in the USAR or Technical Specifications was not increased.
i STP 90-192 l
i TITLE:
Testing of Diesel Generator Fuel Oil Filter Gasket t
DESCRIPTION:
The purpose of this Special Test Procedure (STP) was to install and evaluate the performance of a new gasket material (viton) on the Diesel Generator fuel oil inlet filter. This viton rnaterial was used as the gasket between the bulkhead and canister of the fuel oil system filter.
SAFETY ANALYSIS:
There were no changes in the operation, design, or control of the Diesel Generators System with the performance of this test. The STP installed and evaluated the performance of the viton gasket material in the Diesel generator fuel oil filter. The viton material was verified as being compatible with fuel oil service. No new equipment or components were permanently installed as part of this test procedure and no new safety hazards were created.
STP 90-238 TITLE:
Jet Pump Operability Data DESCRIPTION:
The purpose of this Special Test Procedure (STP) was to collect data which was used in the performance evaluation of the Jet Pump Operability Surveillance when the Recirculation temperature was below rated. Additionally, the data was collected during normal power ascension to improve the repeatability of the present baseline data.
j SAFETY i
ANALYSIS:
This STP did not authorize any equipment alteration or deviations from plant procedures or
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Technical Specifications. The STP defined and collected data during plant heat-up and normal j
startup and did not direct any actions or operations. Therefore, this STP did not affect any plant design bases, design criteria or safety analysis. The only purpose of this STP was to collect and evaluate data. It did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration. Therefore, the affect on overall plant safety was not changed.
STP 90-352 I
TITLE:
MOV Design and Switch Setting Testing j
DESCRIPTION:
This Special Test Procedure (STP) performed in-situ testing of ten (10) Motor Operated Valves (MOVs) to obtain data which will be ut 4 to provide assurance that the MOVs will function when l
subjected to design basis conditions. This STP was implemented as part of the District's MOV Program Project formed in response to the NRC's Generic Letter (GL) 89-10 and Supplements.
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SAFETY ANALYSIS:
This STP was performed with the plant in a cold shutdown condition and was a verification of the design and motor-operator switch settings of the ten valves using normal system operation, design basis accident operating scenarios, and off normal system / plant configurations. Plant equipment was operated in accordance with applicable CNS approved procedures, Technical Specifications, and this STP. All equipment tested was returned to as found position or the position required by the Operations Department.
STP 91-035 TITLE:
Service Water Pump Room Heatup Test DESCRIPTION:
This Special Test Procedure (STP) collected temperature, humidity, and air flow data in the Service Water Pump Room (SWPR) with two service water pumps operating and both HVAC units manually shutdown to simulate a loss of HVAC. During the performance of this STP the SWPR door was opened to determine the affect on the SWPR temperature. The purpose of this STP was to collect data required to demonstrate that the temperature in the SWPR would not exceed the maximum temperature for essential equipment operability with the loss of the non-essential HVAC system. This STP was performed as a commitment to the NRC arising from the Electrical Distribution System Functional inspection (EDSFI) of 1991.
SAFETY ANALYSIS:
This STP took air flow and temperature measurements in the Service Water Pump Room (SWPR) with two service water pumps operating and both HVAC units shutdown. The STP used hand held instruments, thermocouples, and multipoint recorders to measure SWPR air flows and temperature. The measurements were taken at selected locations in the SWPR during operating conditions described above. In addition, temperature in the SWPR was monitored continuously to assure that the area temperature limits were not exceeded. Therefore, operability of all essential and non-essential components was not affected by the performance of this STP. This STP did not increase the probability or consequence of any accidents or malfunctions previously evaluated, nor create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR.
All applicable plant Technical Specifications pertaining to the SWPR and associated non-essential and essential equipment, and room air flow / temperature were monitored at all times.
STP 91-099 and Amendment 1 TITLE:
REC System Flush DESCRIPTION:
This Special Test Procedure provided instructions for performing a velocity flush of the Reactor Equipment Cooling (REC) System. The flush removed any possible corrosion products that may have accumulated in the low flow piping legs of the REC system. This flushing was a result of District commitments in response to Generic Letter 89-13 " Service Water System Problems Affecting Safety-Related Equipment."
SAFETY ANALYSIS:
This Special Test Procedure (STP) did not degrade Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. The REC system was operated in accordance with approved plant procedures 2.2.65
- Reactor Equipment Cooling Water System" and 2.2.65.1
- Reactor Equipment Cooling Water System Valve Checklist" during performance of this STP with the data being collected by utilizing chemistry procedures. Additionally, this STP was performed while the plant was in a shutdown condition thus minimizing required Technical Specification equipment to be declared inoperable during the flush.
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i BTP 91-102 TITLE:
Single Recirculation Loop Drive Flow Correction DESCRIPTION:
The purpose of the Special Test Procedure (STP) was to calculate the correction to drive flow (Delta-W) that is required to obtain effective drive flow when in single recirculation loop operation. The data required to determine this correction was obtained as part of the testing performed for Design Change 90-257 which upgraded the Reactor Recirculation Pumps.
SAFETY ANALYSIS:
This STP only analyzed data that was previously collected to determine a new value of Delta-W for single recirculation loop operation. This STP documented / determined via a calculation a conservative value for the Delta-W correction that can be applied for all core flows including the maximum that can be attained when in single loop operation. This STP did not create an accident or malfunction of a different type, decrease the margin of safety of CNS nor did it directly affect the operation of any plant system.
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REPORTABLE DESIGN CHANGES l
DC 88-169 TITLE:
Modifications of Designated Fume Hoods DESCRIPTION:
This Design Change upgraded three radiological-type fume hoods to protect workers from unencapsulated radioactive materials. This modification consisted of adding fans to the exhaust
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ductwork to increase the air flow and provided HEPA tilters in the exhaust ducts to remove airbome particles from the air prior to releasa.
SAFETY ANALYSIS:
No plant operational requir9ments were affected by this design change. The new fans and filters will improve the performance of the three fume hoods and provide increased personnel i
radiation protection and safety at CNS. Installation of the fans and filters did not create any additional release paths, add radioactive sources, nor change the consequences of a previously analyzed radioactive release. This change had no effect on any defined safety function and did not create an accident or malfunction of a different type than previously evaluated in the USAR.
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i DC 88-201C i
TITLE:
HPCI Room Cooling Modification
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[TSCRIPTION:
This Design Change modified the essential cooling system in the High Pressure Cooling i
Injection (HPCI) Room. This upgrade consisted of replacing the fan coil unit with a larger unit, installing ductwork to provide improved cooling air distribution and, modified the Reactor Equipment Cooling (REC) piping to the new fan coil unit to provide for the increased cooling water flow requirement.
SAFETY ANALYSIS:
The greater cooling capabilities and cooling air distribution brought by this Design Change will decrease room temperature thus increasing the temperature margin for HPCI Room i
1 components during HPCI operation. ' The increase in REC flow and the modifisc. REC piping did not impact the nuclear safety design functions of the REC system. The fan coli unit, its supports, and all ductwork, piping, and components were designed to meet seismic class IS 4
requirements therefore, personnel and nuclear safety were not affected and the change did not create an accident or malfunction of a different type than previously evaluated in the USAR.
DC 88-222F TITLE:
Fire Panel and Miscellaneous Human Engineering Deficiencies (HED's)
DESCRIPTION:
This Design Change was a continuation of the Detailed Control Room Design Review (DCRDR)
Program. This design change modified the existing Cooper Nuclear Station (CNS) Control Room Fire Panel so that the displays and controls were arranged in accordance with estabiished Human Factors Engineering (HFE) guidelines. Application of these HFE guidelines modified the existing control and display positional re!ationships and labeling practices.
SAFETY ANALYSIS:
There were no modifications made to any of the systems that would change the operability or function of the systerr as a result of this Design Change. The majority of the modifications were to panel arrangements and were made to the positional and labeling relationships of the operator controls within the Control Room to HFE guidelines. This modification did not make any functional changes in system operation, components, or equipment. No possibility of an accident or malfunction of a different type than previously evaluated in the USAR or Technical Specifications was created as a result of this change.
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l DC 88-222G l
f TITLE:
Miscellaneous Human Engineering Deficiency (HED) Modifications DESCRIPTION:
This Design Change was a continuation of the Detailed Control Room Design Review (DCRDR) l Program. This design change corrected seven HED's and reorganized the left side of Contro!
j Room Panel 9-5 to provide room for an annunciator CRT and corresponding keypad so that the 4
displays and controis were arranged in accordance with established Human Factors Engineering (HFE) guidelines. Application of these HFE guidelines modified the existing control and display positional relationships and labeling practices.
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ANALYSIS:
There were no modifications made to any of the systems that would change the operability or l
function of the system as a result of this Design Change. The majority of the modifications were l
to panel arrangements are were made to the positional and labeling relationships of the l
operator controls within the Control Room to HFE guidelines. This modification did not make l
any functional changes in system operation, components, or equipment. No possibility of an i
accident or malfunction of a different type than previously evaluated in the USAR or Technical l
Specifications was created as a result of this change.
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i DC 89-049
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TITLE:
Reactor Building Ventilation isolation Radiation Monitor Upgrade.
DESCRIPTION:
This Design Change modified the existing Reactor Building Ventilation isolation Radiation Monitoring System of the Process Radiation Monitorir.g (RMP) System. The modification included the addition of two new sensor converter units and, two new indicator and trip units.
The new components were configured such that each division would have two channels.
SAFETY ANALYSIS:
The system function did not change nor were any setpoints changed with implementation of this design change. Two new setpoints were incorporated for the two new trip units, however, the i
setpoint of the new trip units were set at the same value of the existing trip units. This design change improved the reliability and operability of the Reactor Building Ventilation isolation Radiation Monitoring subsystem by configuring the new and existing units such that each l'
division has two channels. This configuration was reviewed and found acceptable by the NRC in Cooper Nuclear Station License Amendment 147 dated October 10,1991.
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l DC 89-061 TITLE:
Drywell Floor and Equipment Drain Sump Flow Monitor Upgrade DESCRIPTION:
This Design Change replaced the existing components that make up the Drywell Floor Drain Sump, Drywell Equipment Drain Sump, and Reactor Building Equipment Drain Sump Monitoring Systems with advanced technological monitoring equipment. Modifications to the instrument loops included the sump flow monitoring system, sump temperature indication and control systems, and provided essential power to the control and indication circuitry for the Floor and Equipment Drain Isolation Valves.
SAFETY ANALYSIS:
This design change provided newer technological monitoring equipment which, in tum, provides the plant with an overall improved design. Additionally, plant design and reliabliity were improved by providing essential power to the position and indication circuits for the Floor and Equipment Drain isolation Valves. This design change did not in any way, degrade any safety related equipment and/or components and therefore, the ability of these systems to perfom1 their safety function remains unchanged.
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E_89-180A and Amendment 1 TITLE:
Testable Check Valve Actuators (RCIC and HPCI)
DESCRIPTION:
The purpose of this Design Change and Amendment was to improve the testing which verifies the operability of the RCIC and HPCI systems testable check valves. This was accomplished by removing the actuators to further ensure compliance with CNS In-Service Testing (IST) requirements as defined by ASME Section XL The motor operated bypass valves associated with these testable check valves were also removed, in addition, the disc position indicators were modified to improve the valve position indication reliability.
SAFETY ANALYSIS:
The implementation of this Design Change and Amendment was performed while the plant 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 or the affected systems. However the changes did improve the reliability of the RCIC and HPCI testable check valves by modifying the reed switches for more reliable position indication. In addition, the RCIC and HPCI system bypass valves, piping and instrument air drywell penetrations were cut, capped, and hydrostatically (leak) tested. This Design Change did not alter the capabilities of the RCIC or HPCI testable check valves, nor did it change any functions of the affected components curing 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 Changs.
DC 89-215 TITLE:
Control Room Damper AD-1021 A/B Modification DESCRIPTION:
This Design Change (DC) modified the automatic control of Control Room Supply Fan Dampers AD-1021 A/B from fall closed on loss of instrument Air (IA) to manual operation. This Design Change removed the air operators, and capped the IA supply lines. Additionally, this DC eliminated the standby mode of operation for Supply Fans SF-C-1 A/B by replacing the control switches.
SAFETY ANALYSIS:
Changing this system operation to a manual control of the Control Room dampers allows the emergency bypass system to perform its Essential safety function and insures that positive pressure is maintained in the Control Room Envelope.The control switches for the supply fans were replaced which eliminated standby supply fan from auto-starting with the discharge damper closed. Plant design was improved because failure of IA will no longer cause the supply dampers to fall closed and inhibit the positive pressure requirement of the Control Room or interfere with the safety function of the emergency bypass system. Implementation of 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.
DC 89-219 TITLE:
ARTS /ELLLA Implementation DESCRIPTION:
This Design Change (DC) implemented the ARTS /ELLLA modification at Cooper Nuclear Station. ARTS /ELLIA stands for APRM.RBM -Jechnicalfpecification /ExtendedLoadLine Limit Analysis. This DC included the enlargement of the normal power-flow map and the realignme.t of the protective measures (the APRM Rod Block and the Rod Block Monitor). GE report NEDC 21892P contains the analysis that demonstrates reactor operation in the enlarged (ELLLA) portion of the power-flow map is still safe. This DC also reassigned some of the LPRM's to each control rod in the Rod Block Monitor scheme. The new Rod Block Monitor function was changed from flow biased to power biased.
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SAFETY j
ANALYSIS:
Analyses were performed for the operation in the enlarged areas of the power-flow map. These analyses assumed the steady state operation at rated power with various possible core flow conditions. The accidents considered were those considered in the USAR such as loss-of-coolant accidents and the rod drop accident. The review of the accidents and the related safety analyses with the appropriate Technical Specification modifications assured that the 10CFR50.46 and 10CFR100 limits would not be exceeded. Other safety impacts reviewed were the containment responses, reactor internals structural integrity, and the anticipated transient without scram. The anticipated operational transients from load rejection, feedwater controller failure, and rod withdrawal errors were also investigated. These anticipated operational transient analyses established the permissive operating conditions to assure 10CFR20 limits would not be exceeded. This des;gn was reviewed and found acceptable by the NRC in Cooper Nuclear Station License Amendment 151 dated November 29,1991.
DC 89-220 TITLE:
4160/480 Volt Dry-Type Transformer Replacement DESCRIPTION:
This Design Change (DC) provided for the replacement of the existing freon-cooled,4160/480 volt, dry-type transformers with new air-cooled, ventilated,4160/480 volt, dry-type transformers.
The work also involved reconnecting the annunciator cables to the new transformer temperature indicator switches. Additionally, the DC involved the replacement of current transformers (cts) in the 480 volt switchgear ground fault alarm circuits.
SAFETY ANALYSIS:
This Design Change increased the reliability of the 480 volt auxiliary electrical system by replacing the 4160/480 volt, freon-cooled transformers and eliminating future repairs required due to freon leaks, and by replacing the cts which increased the sensitMty of the 480 volt switchgear ground fault circuit. The system performance was not affected by this DC because the new transformers are designed for the same voltage, capacity, temperature rise, and basic impulse level as the existing transformers. Thertfore, the implementation of the DC did not in any way degrade the safety of Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. This DC did not increase the possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the basis for any Technical Specifications.
DC 89-285A and Amendment 1 TITLE:
CRD Check Valve (CRD-CV-25CV and 26CV) Replacement DESCRIPTION:
The purpose of this Design Change and Amendment was to replace the lift check valves with swing check valves. The Y-pattem lift check valves were found to have an excessively high pressure drop across each valve. Amendment 1 removed the valve intemals in both Y-pattem lift check valves and the DC replaced the Y-pattern lift check valves with the swing check valves.
SAFETY ANALYSIS:
The addition of the two essential swing check valves causes a very neg!!@le pressure drop and flow decrease in the CRD system as documented in a NPPD calculatica (NEDC 89-2152). This design does not reduce the flow rate or pressure of the CRD system below pressure and flow requirements stated in the USAR. This design change did not reduce the margin of safety defined in the basis for any Technical Specifications. Speed of control and rod insertion as specified in section 3/4.3 of Technical Specifications remains unchanged. Additionally, implementation of this DC satisfies the concems addressed by GE Potential Reportable Condition (PRC) 89-15 and NRC IN 90-78. The installation of the swing check valves allows little or no leakage outside of secondary containment due to the potential post-LOCA bypass leakage pathway.
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J2GE}AL4 TITLE:
Electric Power to Hazardous Materials Storage Cabinet DESCRIPTION:
This Design Change provided electric power, electric heat, and insulation to the Hazardous Materials Storage Cabinet (HMSC). The Design Change installed a 30 kVA outdoor transformer, outdoor panelboard with associated wiring to the HMSC, and three explosion-proof heaters.
SAFETY ANALYSIS:
This Design Change did not affect any safety related systems, nor did it affect the safe operation or shutdown of any essential system, and was classified as non-essential. This DC did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration. Therefore, the affect on overall plant safety was not changed.
DC 90-035 TITLE:
Radioactive Laundry Upgrade DESCRIPTION:
This Design Change (DC) upgraded the existing radioactive laundry facility to meet the increasing demands placed on it during outages. The new components installed by this design change are more efficient and practical than existing components. This DC also modified the drain line and cold and hot water piping to accommodate the new washer / extractors.
SAFETY ANALYSIS:
This DC was non-essential and did not degrade Cooper Nuclear Station with regard to personnel, equipment, or nuclear safety. This Design Change had no effect on any safety-related system or components and did not affect the availability or reliability of the potable water system, instrument air system, electrical system, or HVAC system. The upgrade to the Radioactive Laundry has no possible impact on any plant safety function or system. The only change is a upgrade of radioactive laundry equipment. All accident analyses remain the same, and no new possible accidents or malfunctions are created.
DC 90-090 Amendment 1 TITLE:
Replacement of SW-MOV-MO89A DESCRIPTION:
This Amendment to Design Change 90-090 was required to remove valves SW-V-1140 and SW-V-1141 from a USAR figure. These valves were connected to the bonnet of the old SW-MOV-MO89A valve by means of a flushing connection. The flushing connections were removed along with replacement of the old SW-MOV-MOB 9A valve by DC 90-090.
SAFETY ANALYSIS:
The flushing connection served no useful purpose and was not intended to be connected to the new SW-MOV-MO89A valve. The function and operation of the valve was not changed nor was the operation of the RHR Service Water Booster System affected. All safety design bases described in the USAR continue to be met. There were no changes in system operating characteristics, or any setpoint changes therefore, no new accidents or malfunctions were created.
DC 90-098 and Amendment 1 TITLE:
Test Jack Installation DESCRIPTION:
This Design Change and Amendment installed test Jacks and keylock switches to facilitate the performance of procedures requiring the use of jumpers. The test Jacks were installed in the Reactor, Control, Turbine, and Diesel Generator buildings. The keylock switches were installed in the Control Room to aid in refueling procedures.
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SAFETY ANALYSIS:
This Design Change improved the methods to bypass and test circuits by installing test Jacks and keylock switches, thereby eliminating the need to install intrusive jumpers, and allowing testing to be performed in a safer, simpler manner. No functional changes were made to the affected systems, and all other design criteria in accordance with app!! cable governing codes, standards, and practices were rnaintained. All margins of safety as defined by the basis for any Technical Specification, USAR, and plant procedures were maintained.
DC 90-112 and Amendment 2 TITLE:
Service Air Cross-Tie Upgrade DESCRIPTION:
This Design Change (DC) and Amendment upgraded the Service Air piping which is cross-tied to various liquid sources. This modification reduced the possibility of liquid leakage back through valves and check valves into the Service Air System. This modification also provided a means to isolate the Service Air system from the liquid sources and provided a path for the liquid to drain from the Service Air System.
SAFETY ANALYSIS:
The Service Air System performance was not affected by this design change. Isolating the Service Air piping from potentially contaminated liquid sources will increase the reliabliity of service air as a source for respirators and reduces the possibility of spreading contamination.
This DC and Amendment only added non-essential components in systems that are not important to safety. This DC and Amendment did not introduce any failure modes that could affect plant operations or nuclear safety. Implementation of this design change did not increase the probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
DC 90-161 TITLE:
Instrument Air Dryer System Upgrade DESCRIPTION:
This Design Change replaced the existing instrument Air dryers and filters with new dual mode dryers (heated / heatless) and filters to offer continuous dry air for plant instruments. This modification also included flow and dewpoint instruments to monitor air flow, and moisture out of the dryers. The air dryers were also provided with power monitoring relays to load shed the heaters upon a power failure so as not to load the heaters onto the Diesel Generators.
SAFETY ANALYSIS:
Modifications performed by 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 changes made by this modification did not create or compound any possible failure modes. The existing failure modes and effects of the system remain appilcable.
All margins of safety defined by the basis of any Technical Specifications, USAR, and plant procedures were maintained. Therefore this DC did not create an unreviewed safety question or have any adverse effect on nuclear safety.
DC 90-168 TITLE:
RWCU High Pressure alarm DESCRIPTION:
This Design Change modified the Reactor Water Cleanup (RWCU) high-low pressure alarm located on board 9-4 in the Control Room. The modification consisted of removing the existing high pressure switch and replacing it with two new high pressure switches to protect the low j
pressure blowdown piping.
)
l 14 1
SAFETY ANALYSIS:
The function of the pressue switches installed by this modification is to protect the RWCU blowdown piping from exposure to pressures greater than its design thereby, increasing the reliability of the protection of thelow pressure blowdown piping. This modification had no affect on the performance of this system. The RWCU serves no safety function as described in the USAR and failure will not directly or indirectly affect any safety-related system, equipment, or components. This DC in no way affected the probability of an accident occurrence, created a previously unidentified accident, or reduced the margin of safety as defined in the basis of any Technical Specification.
DC 90-209 TITLE:
Water Treatment Plant Upgrade DESCRIPTION:
The purpose of this Design Change was to provide a drain line for the water treatment mixed bed effluent. This drain line allows for any impure water to be flushed into the neutralization tank prior to demineralizer train line-up to the demineralized water storage tank.
SAFETY ANALYSIS:
This Design Change did not affect any safety related systems, nor did it affect the safe operation or shutdown of any essential system, and was classified as non-essential. Since none of the steps associated with Design Change 90-209 affected any safe shutdown systems or components and the quality of materials was equal to or greater than those specified in the original construction, implementation of this design change did not increase the probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR. In addition, this design change did not require any changes or additions to the Technical Specifications and involved no reduction in the margin of safety.
DC 90-257 and Amendment 1 TITLE:
RR Pump Upgrade DESCRIPTION:
This Design Change modified the Reactor Recirculation (RR) pump internals (impeller and shaft),
reduced the Control Rod Drive (CRD) flow to the RR pump seals, removed the RR pump suction splitters, modified unused vent / drain lines on the RR pump suction and discharge isolation valves, and provided for refurbishment of the RR pump motors. These changes were implemented in response to technical concems identified in various General Electric Service Information Letters, and Byron Jackson technical alerts and significant event reports primarily related to shaft cracking. The Amendment was written to cover the encapsulation of the vent stub for valve MO53B. The encapsulated pipe consisted of a 2" diameter pipe welded to the valve bonnet and sealed with a blind flange.
SAFETY ANALYSIS:
The pump suction splitters and the CRD seal purge flow to the RR pump seals and RR pump motor refurbishment did not alter the safety function of the RR system. These components are not initiating events for any accident evaluated in the USAR, nor are they required to mitigate the consequences of any accident evaluated in the USAR. Therefore, removal of the splitters, reduction of the seal purge flow, and motor refurbishment did not diminish the ability to maintain and achieve a safe shutdown condition. Modifications to the pump intemals did not degrade pump performance, nor did it affect any safety analysis assumptions Motor refurbishment will enhance motor reliability. The modified vent line configuration represented a direct replacement (with material requirements equal to or greater than original design) which enhanced the structural integrity of the reactor coolant pressure boundary. Therefore, the structural capacity of these components was not diminished. As such, this modification did not increase the possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the basis to any Technical Specification.
15
DC 90-283 Amendment 1 TITLE:
CS-MOV-MO26A/26B and SW-MOV-MO89A/89B Rerate due to Reduced Wall Thickness DESCRIPTION:
This Design Change (DC) and Amendment permanently rerated Core Spray (CS) valves CS-MOV-MO26A and MO26B from ANSI Class 300 to ANSI Class 250 and Service Water (SW) valves SW-MOV-MO89A and MO89B from ANSI Class 300 to ANSI Class 180 in accordance with i
ANSI B16.34,1977 edition. These new ratings are the minimum allowed based on the valves' original design pressure and temperature. The CS valves were rerated due to an as-supplied condition of reduced wall thickness for a portion of CS-MOV-MO26B. The SW valves were j
rerated due to a reduction in valve wall thickness brought on by cavitation / erosion in the valves.
SAFETY ANALYSIS:
This Design Change did not affect the original system design, pressures and temperatures. The original system design pressures and temperatures were used to calculate the new ANSI Class ratings for these valves. The new ratings and lower minimum wall thicknesses do not affect the i
operation or maintenance of these valves in any way. This design change resulted in no physical modifications to the valves or the systems. All previous accident analyses as i
documented in the USAR remain bounding, and no unreviewed safety question was created.
DC 90-292 l
i TITLE:
No Break Power Panel (NBPP) Static inverters DESCRIPTION:
This Design Change replaced the NBPP Static Inverter 1 A with an improved design, completely removed NBPP Static inverter 1B, and changed the power source for four control components in the Reactor Core isolation Cooling (RCIC) System.
SAFETY ANALYSIS:
The modifications 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 affect the function or operation of any system or component related to safe shutdown of the plant. This DC did not create an unreviewed safety question, nor did it reduce the margin of safety defined in the basis to any Technical Specification. Since there was no function or operational changes and the quality of materials were equal to or greater than those specified in the original construction, implementation of this design change did not increase the probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
DC 90-301 TITLE:
DG Exhaust Fan Failure Detection and Screen Wash Pump Gland Seal Alarm Removal DESCRIPTION:
This Design Change provided for the installation of equipment to provide an annunciation alarm upon detection of an essential exhaust fan failure in either Diesel Generator (DG) room. This i
portion of this change is in response to NRC report SSFl 50-298/87-10 in which the NRC stated a concern over the lack of fan failure alarms in the DG rooms. In addition, this Design Change eliminated alarms in the Control Room which indicated low water flow to the gland seals of the screen wash pumps. These alarms were unnecessary and their removal was performed to conform to the annunciator " Blackboard' concept.
SAFETY ANALYSIS:
The addition of the DG exhaust fan failure alarms will alert the operators immediately of this failure and, therefore, provide the operators with more time to respond to this situation. In no way does this change affect the operation of the DGs but will in fact increase the reliability by providing an immediate indication of exhaust fan failure. The removal of the gland seal pump alarms does not affect the function of the screen wash system. This system is non-essential and not safety < elated, but this modification did improve the Control Room annunciator system by removing i nuisance alarm. This DC did not increase the probabifrty of an accident or malfunction, cM not reduce the margin of safety, and did not create a previously unidentified accident or malfu,,ction.
16
_DfdifMD5 and Amendment 1 TITLE:
MINT Project Modification DESCRIPTION:
The Missouri, Iowa, and Nebraska Transmission (MINT) Project was formed so that utilities could jointly construct and operate a 345 kV interconnection between their respective systems.
This Design Change (DC) and Amendment modified the CNS 345 Kv switchyard to accept the additional 345 kV line from Fairport, Missouri, and converted the CNS 345 kV substation from the existing ring bus configuration to a breaker and a half scheme.
SAFETY ANALYSIS:
The implementation of the critical portions of this DC were performed while the plant was in a cold shutdown condition. This DC did not change the original design basis of the offsite electrical system or affect the safety function of the CNS distribution system. However the changes did improve the reliability of the offsite electrical power system by increasing the system grid reliability and stabi!!ty due to the breaker and a half scheme. Furthermore, CNS is analyzed for a complete loss of offsite power and this DC did not after the capabilities of the electricai system, or change any function of the affected system during normal or emergency 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.
DC 90-342 and Amendment 1 TITLE:
RCA Access Facility DESCRIPTION:
The purpose of the Design Change (DC) and Amendment was to install a Radiologically Controlled Area (RCA) Access Facility which provides for a single access / egress location for personnel and equipment to cross the RCA boundary during normal plant operations. This modification is in response to NRC Inspection Report 50-298/90-12 where the NRC identified concems involving the lack of procedures for control of items leaving the RCA. Also, as result of this work, several Post Accident Sampling System (PASS) lines also had to be rerouted but the function of the PASS System was not changed.
SAFETY ANALYSIS:
The modifications performed by this Design Change involved work in the Administrative Building. This building is structurally isolated from Class IS structures and provides no safety function. This modification did not affect nor degrade the performance, or reliability of the PASS; this change however, rerouted the existing system around the RCA. No safety design basis or functional requirements of any systems were affected. Therefore, this modification did not change the existing safety analysis for Cooper Nuclear Station, nor the probability or consequences of an accident as analyzed in the CNS USAR.
j DC 90-351 TITLE:
Reload 14 / Cycle 15 Operation DESCRIPTION:
The purpose of this Design Change was to address the Cycle 15 core reload design and safety analysis. This DC documented the acceptability of the reconfiguration of the core that resulted from the replacement of 164 depleted fuel assemblies with 164 new fuel assemblies. Also included in this Design Change is the replacement of 17 used control blades with 17 new Marathon style control blades. This change also reviewed the results of the safety analysis performed by General Electric for Cooper Nuclear Station Cycle 15 core reload design.
]
i SAFET(
ANALYSIS:
The Cycle 15 reload design was reviewed and analyzed using methodologies described in NEDE-24011-P-A (latest approved version). The analysis for the specified abnormal operational transients and design basis accidents of Section XIV of the CNS USAR remains bounding for 17
the Cycla 15 reload. Therefora, by operating the plint in recordince with the nuclear safety operational requirements as specified in Technical Specifications, the probabl!!ty of occunm.03 or the consequences of an accident or malfunction of equipment important to safety was not increased. No physical changes to plant safety systems were implemented by this Design Change and all safety systems will continue to be operated in their normal (as designed) configuration. Use of the 17 new Marathon control blades was reviewed and found acceptable by the NRC in Cooper Nuclear Station Ucense Amendment 149 dated November 6,1991.
DC 90-360 TITLE:
RPV FW Nozzle Monitoring and HEPA Power Installation DESCRIPTION:
This Design Change provided for the installation of an automated monitoring system for the Reactor Pressure Vessel (RPV) Feedwater (FW) nozzles. The purpose of feedwater nozzle monitoring is to provide the ability to accurately assess the condition of the feedwater nozzles and to ensure their integrity on a continuous basis. The results from nozzle monitoring will be used to justify any changes to future Ultrasonic Testing (UT) and/or dye penetrant testing (PT) inspection intervals. This Design Change also provided for the installation of a permanent power cable for the temporary Drywell High Efficiency Particulate Air (HEPA) filters used during refueling outages.
SAFETY ANALYSIS:
This Design Change provided the ability to continuously assess the condition and integrity of the feedwater nozzles without having to pedorm a periodic PT exam on the nozzles thereby alleviating the need to remove the feedwater spargers on a scheduled basis. Automated UT in conjunction with a continuous monitoring system will perform the function of the PT. As such, this Design Change provides a preferable solution to Unresolved Safety issue A-10, while limiting personnel radiation exposure, possible equipment degradation and challenges to nuclear safety. Personnel radiation exposure is reduced since removal of the feedwater spargers is no longer required on a scheduled basis. Use of an automated UT inspection along with the automated feedwater nozzle monitoring system was found acceptable by the NRC in lieu of performing the PT testing. The installation of the power cable to energize HEPA filters in the Drywell during plant outages in no way impacted any plant system or component. As such, these modifications did not increase the possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the Technical Specifications.
DC 90-381 A i
TITLE:
CRD SDV Relief Valve Removal (CRD-RV-19RV)
DESCRIPTION:
The purpose of this design Change was to remove a Control Rod Drive (CRD) Hydraulic System Scram Discharge Volume (SDV) relief valve CRD-RV-19RV from the north SDV. Removal of this valve was based on General Electric Service Information Letter (SIL) 331, Scram Discharge Volume Design Change Recommendations, and to maintain plant consistency since the relief valve on the south side SDV was also removed based on NRC Information Notice IN 90-18.
SAFETY ANALYSIS:
The removal of the north side SDV relief valve did not affect system performance of the SDV or the CRD system. The SDV is still able to contain all of the water discharged after a scram and scram insertion times were not affected. This relief vatve is not required to perform any re!ief function for the SDV and the scram discharge piping was not adversely affected by remova! of this valve. The safety function of the SDV was unchanged, and the integrity of the SDV piping was not degraded by this modification, the GE SIL documents that the SDV piping was analyzed for occasional exposure to RPV design pressures. As such, this modification did not increase the possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the basis for any Technical Specification.
18
DC 94-381B TITLE:
CRD SDV Relief Valve Removal (CRD-RV-12RV)
DESCRIPTION:
The purpose of this design Change was to remove a Control Rod Drive (CRD) Hydraulic System Scram Discharge Volume (SDV) relief valve CRD-RV-12RV from the south SDV. Removal of this valve was based on NRC Information Notice IN 90-18, and General Electric Service Information Letter (SIL) 331, Scram Discharge Volume Design Change Recommendations.
SAFETY ANALYSIS:
The removal of the south side SDV relief valve did not affect system performance of the SDV l
or the CRD system. The SDV is still able to contain all of the water discharDed after a scram L
and scram insertion times were not affected. This relief valve is not required to perform any relief function for the SDV and the scram discharge piping was not adversely affected by j
removal of this valve. The safety function of the SDV was unchanged, and the integrity of tfm i
SDV piping was not degraded by this modification, the GE SIL documents that the SDV piping j
was analyzed for occasional exposure to RPV design pressures. As such, this modification did not increase the possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the basis for any Technical Specification.
(
I DC 91-123 c
TITLE:
Isolation of Service Water to DG H&V Units DESCRIPTION:
This Design Change permanently isolated the Service Water (SW) to the Diesel Generator (DG)
Heating & Ventilation (H&V) units. Service Water to these units was determined not to be required for either normal or emergency operation of the DGs under design basis conditions.
Isolating the SW to these units will eliminate routine and periodic maintenance requirements for the cooling coils.
SAFETY ANALYSIS:
Implementation of this design change did not affect the operability of the DGs at design basis temperatures or conditions.
The DG room temperatures remain under the maximum
[
temperature limits, for operability of the essential electrical equipment located in the DG rooms a
with no Service Water flow to the H&V units. The air flow through the H&V units will maintain DG room temperature under the maximum temperature limits. There were no changes in DG system operating characteristics, therefore, no affect on overall plant safety.
i ESC 91-133 l
TITLE:
Replacement of RPV Closure Botting Material DESCRIPTION:
This Equipment Specification Change (ESC) covered the replacement of 48 Reactor Pressure i
Vessel (RPV) closure bolting components (closure studs and bottom closing screws) with components obtained from Shoreham Nuclear Station. All original CNS closure nuts and closure washers were reinstalled. The original closure studs and the bottom closing screws from Shoreham were manufactured from commercial alloy steel and were verified interchangeable with those at CNS.
SAFETY ANALYSIS:
The replacement closure studs and bottom closing screws (ASTM A540 B23, Class 3) have the same form, fit, and function as the original components (ASTM AS40 B24, Class 3). The material change from ASTM AS40 B24, Class 3 to ASTM A540 B23, Class 3 did not affect the mechanical properties of the botting material (i.e. yield strength, tensile strength, and impact test values), including allowable stress values. The safety design basis and functional requirements of the closure botting material were not affec;ed, since the critical characteristics of the replacement material meets the requirements of the original closure botting material.
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REPORTABLE ACTIVITIES i
l SDC 90-016 TITLE:
Emergency Response Data System (ERDS)
DESCRIPTION:
This Software Design Change (SDC) provided Cooper Nuclear Station the ability to communicate with, and transmit plant performance and environmental data to the NRC during an Alert or higher emergency classification level in accordance with the CNS emergency plan.
SAFETY ANALYSIS:
The ERDS software runs on the Plant Management Information System (PMIS) which is an isolated, non-safety-related system, thus, no probability exists for any safety related equipment being affected by this SDC. The PMIS computer system is part of the plant monitoring system only, and does not directly affect the performance or operation of any plant system including those important to safety. Therefore, performance of this SDC did not involve an unreviewed safety question.
Procedure Chance Notice (PCN) 2.2.20.2 (Revision 3)
TITLE:
Operation of Diesel Generators (DG) from Diesel Generator Room DESCRIPTION:
The purpose of this procedure change was to add protective steps to pull the relay paddles on the DG #2 differential protection relays to prevent these relays from actuating and causing DG
- 2 to trip and lockout. This event could only happen during an Appendix R scenarlo in fire zone 8A (Aux Relay Room) and fire zone 9 (Cable Spreading Room).
SAFETY ANALYSIS:
The removal of the DG #2 differential relay paddies during Appendix R fire scenarios does not increase the damage potential of the DG #2 and associated power cable. The Appendix R scenario requirements do not postulate any damage or failures other than those caused by the fire, therefore, the removal of the DG #2 differential protective relay paddles during an Appendix R scenario only removes a potential fire-induced trip signal to DG #2 with no additional risk to DG #2. The DG #2 relay paddles will only be disabled during an A!!emate Shutdown fire scenario as described in the basis for the Technical Specification and will not affect the margin of safety.
Procedure Chance Notice (PCN) 2.2 24 (Revision 22)
TITLE:
250 Volt DC Electrical System i
DESCRIPTION:
The purpose of this Procedure Change Notice (PCN) was to allow the use of a fire watch as a compensatory measure to address Appendix R concems when the 250 volt *C* battery charger is used to feed "B" Bus. Additionally, this PCN documented the acceptability of raising the 250 volt DC float voltage from 262 volts to 270 volts as per vendor recommendations.
SAFETY ANALYSIS:
The use of the "C" battery charger as a spare for either the A or B charger for the 250 volt DC system does not increase the possibility of a loss of DC electric power since the "C" charger meets the same design and performance criteria as the other chargers. Therefore, the use of the "C" 250 volt battery charger during power operation cannot increase the probability of an accident described in the USAR. The increase in float voltage to the battery was suggested by the vendor to aid in the prevention of corrosion bulidup on the battery plates. This float vo!! age increase in no way impacted any system or component operation. Additionally, the use of *C*
250 volt battery charger during power operation, in conjunction with the compensatory fire watches for Appendix R concems does not create the possibility of an accident not described in the USAR.
20
P_rncedure Chance Notice (PCN) 12.25 (Revision 23) i TITLE:
125 Volt DC Electrical System DESCRIPTION:
The purpose of this Procedure Change Notice (PCN) was to allow the use of a fire watch as a compensatory measure to address Appendix R concems when the 125 voit "C" battery charger is used to feed *B" Bus. Additionally, this PCN documented the acceptability of raising the 125 voit DC float voltage from 129 volts to 130 volts as per vendor recommendations.
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SAFETY ANALYSIS:
The use of the "C" battery charger as a spare for either the A or B charger for the 125 volt DC l
system does not increase the possibility of a loss of DC electric power since the *C" charger l
meets the same design and performance criteria as the other chargers. Therefore, the use of the "C" 125 voit battery charger during power operation cannot increase the probability of an accident described in the USAR. The increase in float voltage to the battery was suggested by the vendor to aid in the prevention of corrosion buildup on the battery plates. This float voltage increase in no way impacted any system or component operation. Additionally, the use of *C*
l 125 voit battery charger during power operation, in conjunction with the compensatory fire l
watches for Appendix R concems does not create the possibility of an accident not described in the USAR.
i Procedure Chance Notice (PCN) 5 4 3.1. (Revision 9)
TITLE:
Post-Fire Operational Information i
DESCRIPTION:
This Procedure Change Notice (PCN) provided a method for the operators to restore control of the Diesel Generator fuel oil transfer pumps after a loss of control power to the transfer pumps as a result of an Appendix R fire scenario in certain locations at CNS.
SAFETY ANALYSIS:
This PCN only provided operators tho ability to manually control the transfer of diesel fuel oil in the event of cont. M cable failure as a result of fire. Operation of the fuel oil transfer pumps during normal operation was not changed. This PCN involved no technical or operational aspects that directly affect normal station operation. This procedure change did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration.
i Procedure Chance Notice (PCN) EPIP 5.7 8 (Revision 12)
TITLE:
Activation of OSC DESCRIPTION:
The Purpose of this Emergency Plan implementing Procedure (EPIP), Procedure Change Notice
[
(PCN) was to combine the Chemistry / Health Physics, Maintenance, and instrument &
Control / Electrical Operations Support Centers (OSCs) into one OSC. Also included in this PCN was discussion of the new OSC staffing and location, and a new accountability sign-in.
SAFETY i
ANALYSIS:
Consolidating the three OSCs into one OSC does not in any way affect plant safety or response. The three craft disciplines identified in the this procedure will continue to respond l
to an emergency as outlined in the procedure. Furthermore, emergency response procedures cuch as this do not increase the probability of an accident described in the USAR, nor create the possibility of an accident not described in the USAR, and do not reduce the margin of safety, because the procedure provides instructions for coping with/ mitigating an emergency i
(accident) that may have already occurred.
21 1
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. _P oc'edure Chance Notice (PCN) EPIP 5.7.10 (Revision 16) l TITLE:
Personnel Assembly and Accountability DESCRIPTION:
The Purpose of this Emergency Plan implementing Procedure (EPIP), Procedure Chage Notice (PCN) was to combine the Chemistry / Health Physics, Maintenance, and Innrument &
Control / Electrical Operations Support Centers (OSCs) into one OSC as described in this procedure. Also included in this PCN was a description of the new designated assembly area.
SAFETY l
j ANALYSIS:
Consolidating the three OSCs into one OSC does not in any way affect plant safety or response. The three craft disciplines identified in the this procedure will continue to respond -
to an emergency as outlined in the procedure. Furthermore, emergency response procedures such as this do not increase the probability of an accident described in the USAR, nor create l
the possibility of an accident not described in the USAR, and do not reduce the margin of safety, because the procedure provides instructions for coping with/ mitigating an emergency i
(accident) that may have already occurred.
Procedure Chance Notice (PCN) EPIP 5.7.15 (Revision 11) l TITLE:
Repair /Re-Entry or Rescue DESCRIPTION:
The Purpose of this Emergency Plan implementing Procedure (EPIP), Procedure Change Notice (PCN) was to combine the Chemistry / Health Physics, Maintenance, and instrument &
Control / Electrical Operations Support Centers (OSCs) into one OSC. Also included in this PCN was discussion of the new location of the OSC Rescue tool cabinet.
t i
SAFETY
[
ANALYSIS:
Consolidating the three OSCs into one OSC does not in any way affect plant safety or response. The three craft disciplines identified in the this procedure will continue to respond l
to an emergency as outlined in the procedure. Furthermore, emergency response procedures j
such as this do not increase the probability of an accident described in the USAR, nor create -
the possibility of an accident not described in the USAR, and do not reduce the margin of safety, because the procedure provides instructions for coping with/ mitigating an emergency (accident) that may have already occurred.
i l
Procedure Chance Notice (PCN) EPIP 5.7.22 (Revision 14) i TITLE:
Communications DESCRIPTION:
The Purpose of this Emergency Plan implementing Procedure (EPIP), Procedure Change Notice (PCN) was to combine the Chemistry / Health Physics, Maintenance, and instrument &
l Control / Electrical Operations Support Centers (OSCs) into one OSC as described in this procedure. Additionally, this PCN discussed the new Bonephone locations and numbers.
SAFE 1Y ANALYSIS:
Consolidating the three OSCs into one OSC does not in any way affect plant safety or response. The three craft disciplines identified in the this procedure will continue to respond j
to an emergency as outlined in the procedure. Furthermore, emergency response procedures such as this do not increase the probability of an accident described in the USAR, nor create the possibility of an accident not described in the USAR, and do not reduce the margin of safety, because the procedure provides instructions for coping with/ mitigating an emergency i
(accident) that may have already occurred.
i 22 l
l
-.- 2
Oth#r ActMties TITLE:
Evaluation of Maximum Safe Flood Depths (MSFD) for Cooper Nuclear Station DESCRIPTION:
The purpose of this evaluation was to analyze the safety implications of a change in the MSFD for CNS as listed in the Updated Safety Analysis Report (USAR). This evaluation was performed in response to recommendation 1 of INPO Significant Operating Event Report (SOER) 85-5.
SAFETY ANALYSIS:
The original licensing bases for intemal plant flooding required consideration of the failure of seismic class IIS piping only. The above evaluation went beyond this bases and analyzed flooding as a result of MELBs and HELBs regardless of seismic classification. The analysis showed that safe shutdown of the plant is not impaired by changes in the MSFDs at CNS. The possibility of a accident previously evaluated in the USAR is not created because failure of seismic class IS piping is currently contained in the USAR. The margin of safety as defined in the basis for any Technical Specifications is not reduced.
Other Activities TITLE:
Evaluation of Reactor Building Exhaust Ventilation High Radiation Signal DESCRIPTION:
The purpose of this safety evaluation was to correct a discrepancy between the Primary Containment isolation System (PCIS) section of USAR chapter Vil and the Technical Specifications concoming the Reactor Building Exhaust Ventilation High Radiation Signal (Group 6 isolation). The Technical Specifications list this as a primary containment isolation signal that closes primary containment isolation valves (purge and vent valves). However, in USAR Chapter Vil, it is not included in the list of containment isolation signals. To make the USAR consistent with the Technical Specifications, the high radiation signal will be identified as a containment isolation signal, and a section describing this signal will be added to the USAR in 1993.
SAFETY ANALYSIS:
The Reactor Building Exhaust Ventilation High Radiation Signal has always been a primary containment isolation signal (close signal for the purge and vent valves) however, the original CNS USAR and subsequent revisions failed to discuss this signal in chapter Vll. This change only added the description of The Reactor Building Exhaust Ventilation High Radiation signal as a primary containment isolation signal to USAR Chapter Vll. There are no Systems, Subsystems, Components, or Computer Software affected by this change. Addition of the PCIS signal that actuates primary containment closure to the USAR does not create a possibility for a accident or malfunction of a different type than any previously evaluated in the USAR. This signal has always been present at CNS however, discussion of this signal was not included in the USAR. The only change is a description of the signal in the USAR, all accident analysis remain the same, and no new accidents or malfunctions are created.
Other ActMties TITLE:
Nuclear Power Group (NPG) Organizational Changes DESCRIPTION:
This activity documented the NPG organizational changes that were modified by the creation of the Site Manager position and a Plant Manager position. Additionally, the NPG organization was modified by shifting the reporting responsibilities of some individuals, and various position title changes.
SAFETY ANALYSIS:
Changes in the NPG organization structure do not affect the design or operation of any plant system, structure, or component described in the USAR accident analysis. The organization 23
changes that took plac3 cffect only which position will perform the required responsibilities. All persons filling the new positions were qualified to perform the assigned tasks and responsibilities. These changes are considered to be an administrative change to the NPG organization which does not affect the performance of the organization to effectively respond to plant transients or emergencies. For additional information reference CNS Proposed Technical Specification Change No.111 submitted to the NRC on October 8,1992.
Other Activities TITLE:
Evaluation of Guidelines for Preparation of Emergency Operating Procedure Flow Charts DESCRIPTION:
This evaluation was conducted to ensure that Revision 3 of the Guideline for Preparation of Emergancy Operating Procedure Flow Charts is consistent with Revision 1 of the CNS Emergency Procedure Guideline (EPG), Revision 4 of the Bolling Water Reactor Owners Group (BWROG) EPG, and did not conflict with either the design basis or operation of CNS as described in the USAR and other licensing documents.
SAFETY ANALYSIS:
The implementation of the revised Flow Chart Guideline (FCG) did not modify the design basis of the plant, the flow charts are only used after plant conditions have reached EOP entry conditions. Therefore, the revised FCG cannot increase the probability of occurrence of any event analyzed in the USAR since the flow charts are only used after an event has commenced.
Additionally, the possibility of accidents or malfunctions of different types than those listed in the USAR was not created.
i 24
111.
PERSONNEL AND MAN-REM EXPOSURE 25
PERSONNEL AND MAN-REM GY WORK AND JOB FUNCTION NUMBER OF PERSONNEL
( > 100 mrem )
TOTAL MAN-REM Station Utility Contractor a Station Utility Contractor &
WORK AND JOB FUNCTION Erroloyees Employees Others Employees Employees Others REACTOR OPERATIONS & SUPV.
Maintenance Personnel 3
0 1
0.178 0.000 0.006 Operating Personnel 32 0
0 8.343 0.000 0.000 Health Physics Personnel 17 0
9 2.029 0.000 0.676 Supenrisory Personnel 2
0 0
0.143 0.000 0.000 Engineering Pusonnel 4
2 4
0.843 0.121 0.019 ROUTINE MAINTENANCE Maintenance Personnel 58 0
10 17.904 0.000 3.318 Operating Personr.el 6
0 0
0.084 0.000 0.000 Health Physics Personnel 27 0
14 15.756 0.000 5.805 Supervisory Personnel 0
0 0
0.000 0.000 0.000 Engineering Personnel 2
11 3
0.008 2.08' 2.314 SPECIAL MAINTENANCE Maintenance Personnel 1
0 0
0.018 0.000 0.000 i
Operating Personne!
5 0
0 0.430 0.000 0.000 Health Physics Personnel 2
0 0
0.119 0.000 0.000 Supervisory Personnel 1
0 0
0.132 0.000 0.000 Engineering Personnel 0
0 0
0.000 0.000 0.000 WASTE PROCESSING Olnieriance Personnel 2
0 0
0.026 0.000 0.000 Operating Personnel 4
0 0
1.894 0.000 0.000 l
Health Physics Personnel 4
0 2
1.547 0.000 0.104 Supervisorf 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 0
0 0
0.000 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 0
0.000 0.000 0.000 INSERVICE INSPECTION l
Maintenance Personnel 0
0 1
0.000 0.000 0.264 Operating Personnel 1
0 0
0.025 0.000 0.000 Health Physics Personnel 0
0 0
0.000 0.000 0.000 Supervisory Personnel 1
0 0
0.048 0.000 0.000 Engineering Personnel 0
0 0
0.000 0.000 0.000 TOTAL Maintenance Personnel 59 0
11 18.126 0.000 3.588 Operating Personnel 36 0
0 10.776 0.000 0.000 Health Physics Personnel 29 0
14 19.451 0.000 6.585 Supervisory Personnel 2
0 0
0.323 0.000 0.000 Engineering Personnel 4
11 3
0.851 2.205 2.333 GRAND TOTALS 130 11 28 49.527 2.205 12.506 26 l