ML20106D066
| ML20106D066 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/31/1992 |
| From: | ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20106C643 | List: |
| References | |
| NUDOCS 9210070164 | |
| Download: ML20106D066 (58) | |
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l CLINTON POWER STATION 10CFR50.59 REPORT FOR MAY 1991 TilROUCll NAY 1992 9210070164 920924 PDR ADOCK 05000461 Y
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DEFINITION OF ACRONYMS FOR DOCUMENTS EVALUATED ACN Advance Change Notice ASME American Society of Mechanical Engineers ASTM -
American Society of Testing and Materials CR Condition Report CPS Clinton Power Station ECN Engineering Change Notice EPIP -
Emergency Plan Implementing Procedures FA Field Alteration FECN -
Field Engineering Cha;go Notice HVAC -
Heating, Ventilation, and Air Conditioning IEEE -
Institute of Electrical and Electronic Engineers, Inc.
NFPA -
National Fire Protection Association PDR Procedure Deviation for Revision P&ID -
Piping and Instrumentation Diagram TM Temporary Modification TPD Temporary Procedure Deviation USAR -
Updated Safety Analysis Report S
Supplement R
Revision l
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CLINTON POWER STATION 10CFR50.59 REPORT FOR l
l MODIFICATIONS FROM MAY 1991 THROUGH MAY 1992 l
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PAGE 2 INSTALL FLANCES IN AUXILIARY STEAM SYSTEM Docu*ent Evaluated: FA A5F017 Log Nunber: 91 0030 Flanges were installed between the auxiliary steam reboilers and their associated condensate drain tanks to facilitate removin6 the reboilers from service for maintenance.
This modification was released for operation for the "A"
reboiler only.
The installed flanges do not affect operation or performance of the associated components.
The change adds no new failure modes as other piping connected to the reboilers is also flanged.
REPLACE OBSOLETE AUTOMATIC RECIRCULATION VALVE VITH A NEW MODEL IN THE AUXILIARY STEAM SYSTEM Docunent Evaluated: FA ASF018 Log Nmber: 91 0088 The electrode boiler feed pumps in the auxiliary steam (AS) system are provided with minimum flow valves.
The present low pressure automatic recirculation control (LARC) valves are obsolete and were replaced with automatic recirculation control (ARC) valves.
This is a change to the facility as described in USAR Figures 3.6-1 and 9.5-6.
The function and sizing of the ARC valve is the same as the LARC valve. No malfunctions of the AS system are evaluated in the USAR.
PROVIDE CAPABILITY TO ADJUST C00LINC WATER FLOW TO UPPER AND LOVER BEARINCS OF THE REACTOR RECIRCULATION PUMP MOTORS Docwent Evaluated: FA ttF010 Log Nwber: 92 0009 The reactor recirculation pumps provide forced circulation of coolant through the reactor pressure vessel core through jet pumps integral to the reactor vessel.
Each reactor recirculation (RR) pump motor has an upper and lower bearing provided with coolers.
Previously, there was no means for balancing the cooling water flow between the upper and lower bearings. An investigation showed that the cooling water flow through the coolers exceeded the maximum-design flows, This change provided a flow meter and a flow control valve at the outlet of each of the upper and lower bearing coolers on both of the RR pump motors.
This change required a revision to.USAR Figures 3.6 1 and 9.2-3 and the description of the RR system. Neither the RR pump motors nor the component cooling water system are required to operate during design basis accidents.
This change will extend the operating life of the coolers by.
-providing means to control flow through them.
No new type of failure has been introduced in'the USaR, nor has there been an increase in the consequences or potential.for a failure previously evaluated, i
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PAGE 3 UPDATE USAR TO REFLECT CHANCES TO OPERATION OF MSIV LEAKAGE CONTROL SYCTEM INBOARD AND OUTBOARD ROOM COOLERS Document Evaluated: FA C-F054, FECN 24378 Log Nueer: 90 0133 Previous-design changes required the Main Steam Isolation Valve (MSIV) Leakage Control System (LCS) inboard and outboard rooms, high pressure core spray-(HPCS) system room, and combustible gas control system (CGCS) equipment room coolers to operate continuously during normal plant operating conditions.
The temperature switches which started the room coolers were not qualified and were subsequently deleted.
This change deletes drawing notes which indicate that the temperature switches will be replaced.
It was determined that the respective room coolers need not run continuously during normal plant conditions.
The MSIV LCS inboard and outboard room temperatures werc calculated to increase from 104*F to 108'F and 115'F respectively.
The HPCS and CGCS room temperatures were determined to be within design limits.
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equipment qualification packages were revised.
Where the qualified life of equipment has been reduced, the affected equipment will be replaced at a higher frequency.
REPLACE THE DIVISION I BATTERY Document Evaluated: FA DCF004 Log Nunber: 91-0124 & R1 Battery 1DC01E provides a reliable backup source of DC power for Division 1 equipment. Due to design of the existing battery and its limited design margin, it was necessary to increase the capacity of the battery.
The existing battery was rated at 1138 ampere-hrs.
The new battery is rated at 1708 ampere-hrs. The-design change was reviewed for electrical' load impact, structural adequacy, room temperature, hydrogen evolution, room access, fire protection, the effects of voltage drops and short circuit currents on existing components, relay settings and circuit breaker replacement.
The new battery has cast lead battery posts with a copper insert, to provide structural _ strength, which the previous battery did not have.
Problems have been encountered in the past with electrolyte attacking the copper insert causing a decline in cell voltage. New manufacturing techniques have corrected this problem.
It is judged that the probabilities of a malfunction involving a decline in cell voltage have not increased.
A revision to the safety evaluation allowed a partial release on the Division 1 nuclear system protection system inverter. The inverter was to be placed back in service and supplied power from the-Division 1 battery charger IDC06E, via Division 1 125-volt DC bus IDC13E, while the Division 1 battery was being replaced.
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PAGE 4 CllANGES TO EMERGENCY DIESEL GENERATOR DOCUMENTA'IION Doewent Evaluated: F A DGF0?6 Log thirber: 91 3100 Each diesel generat.or (DG) contains an expansion tank for cooling water which is rated for 50 psi.
The expansion tank provides a reservoir for cooling water a s it expands.
Each expansion tank has a level switch which will ener6 ze an annunciater at a predetermined low level.
The expansion tank also 1
contains a pressure er which 311ows filling of the expansion tank and pressure release betv an 6.5 and 8 psi.
The low level annunciator is nnt affecte3. A condition report iden'ified that the exparwton tank 1cvel switches for thn Division I and II DGs (1LS-DC285, 256, 287, 288) were not in the seismic qualification (Sn) program even O agh they are seismic category I.
During the condition report investigation, it was found that the r:bove level switches were connected to safety-related ll; buses with no supporting documentation.
Field Alteration DGF026 Implemented the following changes:
- 1. Allows non-lE Level St(tches ILS-DG285, 286, 287, 288 to be connected to 1E busen.
- 2. The expansion tank level switch for the Division 111 DG is required as a pressure boundary to prevent draf ning of the Division 1;l DG cxpansion tank.
The switch already meets the requirunents for the classi ficat ion change.
- 3. For Division I and 11 DGs, the design maximum p assure for the expansion tank level switches was revised from 50 psi to 16 psi.
Since the fili cap for the expansion tank relieves pressure between 6.5 and 8 psi, this change will allow a 100% margin in the pressure rating of the switches.
- 4. USAR Table 9.5-13 was revised to include the Division I and Division II level switches, ILS-Da285, 286, 287, 288.
- 5. USAR Table 9.5-13 will delete the Maintenance and Test Frequency column.
Currently, the USAR Tabic 9.5-13 Maintenance and Test Frequency column indicates " Test frequencies will not exceed one year" The column is no longer necessary because IP's Diesel Generator Qualification Program delineates the preventive maintenance requirements for the diesels.
This evalcation indicates that these changes do not increase the probability or consequences of a malfunction of the diesel generators nor do they create a new type of malfunctlon.
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PACE 5 CLOSE ISOLATION VALVE ON CO2 FIRE SUPPRESSION STORAGE TANK l
Document (voluated: FA FPF011 Log Nwtwr: 91-0011 ll l
A carbon dioxide (CO ) fire _ suppression system is provided for the three 2
emergency diesel generator rooms.
The system design conforms with National Fire Protection Association (NFPA) Standard NFPA_12 "CO2 txtinguishing Systems".
There are two CO2 storage tanks connected in parallel to the main i
discharge header.
This field alteration changes the normal position of the isolation valve on one of the tanks from normally open to normally closed.
This' change also adds a chain wheel operator so that the normally closed valve can be opened quickly.
Testing has shown that with the isolation valve closed the system provides sufficient quantity and concentration of CO -
2 DEACTIVATE MAIN CONTROL ROOM FIRE PROTECTION MONITORING OF THE BLUE WAREHOUSE Doewent Evaluated: FA FPF028 Log Norber: 91 0084 This modification deactivates the main control room monitoring of the fire protection annunciation and alarm logging of the blue warehouse fire protection system.
The change was made to remove the monitoring of the warehouse fire protection from the responsibility of the control room operators, and place it with site security similar to the other facilities outside of the protected area. The responsibility of monitoring the blue warehouse is placed where it can be more directly handled. The change in the fire protection monitoring responsibility will not affect any of the plant nuclear safe ty-related sys tems.
The desi,n change involves minor wiring changes to delete alarms and to eliminate inputs to the alarm circuits.
DELETE NOTE FROM USAR FIGURE REQUIRING SPECIFIC CHECK VALVE Docunent Evaluated: FA FPF031 Log Nuter: 91 0064 l
l A change was made to drawing M05-1039, sheet 6 (same as USAR Figure 9.5-1, Sheet 6, Note 6).
This change was to delete the requirement that a specific model of backflow preventer be-used on the discharge side of the jockey fire pump. This change allows the use of other equivalent types.
This change was made because a condition report identified that the existing backflow l
preventer did not match the specification.
The existing model is equivalent the model specified by the note.
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p PAGE 6 PROVIDE EMERGENCY POWER TO CHEMISTRY-LA'>0RATORY Document Evaluated: FA LLf006 -
Log huriber: 90 01?'
This_ field alteration installed emergency power to receptacles in the chemistry laboratory which are utilized for post accident sample analysis following a. loss of coolant accident (LDCA).
This-change makes it possible for sample analysis following a LOCA that is accompanied by a toss of offsite power (LOOP).
The specific change entailed connecting the transformer which provides power to these receptacles to the Division II diesel generator motor control center (MCC) IB.
Since the receptacle circuit is not Class lE, the connection is being made through acceptable isolation devices. This additional load was evaluated by revising the diesel Senerator load calculation.
During this revision, it was determined that, due to more recent input from the respective vendors and the mathematical corrections, the total loading on all three diesel generators has been decreased.
The total loading does not exceed the name plate rating.
Fuel oil storage requirements per technical specifications are not affected.
INSTALLATION OF GE8B FUEL BUNDLES Docunent Evaluated: FA NBF010 Log N mber; 91*0115 Field alteration NBF010 evaluated the use of the GE8B fuel design.
GE6 and GE7B fuel designs are currently used.
CE8B fuel has a higher helium backfill pressure which allows a higher maximum linear heat generation rate (MLHGR)
(14.4 kilowatt / foot [kw/ft] vs. 13.4 kw/ft) and allows higher batch average burnup (38000 megawatt-day / metric ton [ Mwd /MT] vs, 33000 Mwd /MT).
In addition, GE8B fuel has a gadolinia-rich zone near the top of the bundle, which improves reactivity shutdown margin.
GE8B fuel has been generically reviewed by the NRG and has the same form, fit and function as GE6 and GE7B fuels for reload design.
In addition, over 5000 GE8B fuel bundles have.been-used in other reactors. The reason for the' change is that GE8B improves the margin to MLHGR, improves cold shutdown, and reduces fuel costs.-
The consequences of accidents are not altered because GE88, GE6, and GE7B fuels have the same design basis.
NRC-approved methods were used to evaluate: MLHGR, minimum critical power ratio (MCPR), maximum average plant linear heat generation rate (HAPLHGR) before each. cycle for normal operation; anticipated operational occarrencesi-and loss of coolant accidents (LOCA) events to ensure - that there-is no increased fuel failure.
In addition, cold shutdown and reactor stability are evaluated each cycle to ensure that there is no effect on fuel failures.
Regarding defective fuel during normal cperations, based-on historical data, the GE8B fuel pin failure rate is 0.99999 which is an improvement over the CE7B fuel pin failure-rate of 0.99994.
Therefore, the consequences of=
failures are not -changed.
Regarding the fuel bundle drop accident, GE8B fuel has the same number of fuct rods, the same uranium mass, and the same design bacis as CE6 and GE7B fuels.
Therefore, this design change does not alter the consequences of a fuel bundle drop accident.
The basis for the conclusion is that the bundle bur,,up -15
,i-PACE 7 consistent with the burnup used for the Updated Safety Analysis Report, Section 12.2.1, and the associated source term for the bundle.
_Regarding an anticipated transient without_ scram (ATVS) event, there is an additional consideration that has to do with Emergency Guidelines, Appendix C.
The change affects the maximum suberitical banked withdrawal position (MSBWP).
The_MSBVP is "00" for GE8B fuel compared to "02" for GE6 and GE7B fuels.
This change ensures that-there is no effect on fuel failure if there is a potential for an ATWS and the consequences are not altered.
Maximum core uncovery time limit (MCUTL) and minimum alternate reactor pressure vessel pressure (MARFP) in the Emergency Procedure Guidelines, Appendix C are affected because MLHCR increases -from 13.4 kw/ft to 14.4 kw/ft. When multiple failures were considered, new MCUTL and MARFP were calculated to ensure that there is no effect on fuel failures during extraordinary events and the consequences of failures are not altered.
Regarding effects on fuel failures for very high bundle burnup, the evaluation concluded that there was no significant effect except for increased iodine levels for the fuel bundle drop accident.
However, the batch average discharge burnup will not approach the burnup value of 29,200 Mwd /MT which is used in the analysis until the end of Cycle 5.
Assessments are planned in order to determine the effect of increased burn-up capability of GE8B fuel on the source term for future cycles.
Because the helium backfill pressure in the fuel pin is higher than in the old designs, cladding stress at system pressure is reduced and heat transfer from the pellets and out of the pins is improved Improved heat transfer will decrease fission gas release to the gap between the fuel pellet and the cladding which lowers peak pin pressure during operation.
This permits the MLHGR to be increased to 14.4 kw/ft without exceeding the 1% cladding strain safety limit.
This design change has no effect on the bases of the technical specifications. However, as part of the reload design, the Core Operating Limits Report will be changed to reflect 14.4 kw/ft for the GE8B fuel, REPLACE RCIC TURBINE EXHAUST FLANGED JOINT WITH A E LDED CONNECTION Docunent Evaluated: FA RIF010 Log Nunber: 91-0110 The reactor core isolation cooling (RCIC) system turbine exhaust line contained a flanged connection that would be exposed to containment atmosphere following drawdown of_the suppression pool after a design basis loss of cooling accident. Corrective actions included an Appendix J, Type B leak rate test during the recond refueling outage and replacement of the flange'with welded pipe during the third refueling cutage.
This change eliminated the flanged connection, which eliminates the possible leakage path and the need for Appendix J testing of thir penetration. The piping modificatftn design conforms to the design criteria which is ASME Section III, Class 2, seismic category 1.
This change does not affect RCIC system operation.
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PAGE 8 EXPAND CONFIGURATION OF THE SERVICE AIR SYSTEM Docunent Evaluated: FA SAF006 Log Nunber: 90 0117 This field alteration expands the service air (SA) system to include existing piping installed during the early construction phase of the plant.
This piping will be used to provide an alternate path for air to be supplied to the SA header from an outside source.
This-equipment is-upstream of the instrument air system.
The design of the plant is such that a failure of SA will not compromise any nuclear safety-related system or component and will ot prevent the safe shutdown of the reactor.
INCREASE THE LOV POWER SETPOINT PRESSURE Document Evaluated: FA TGF008 53 Log Neber: 91-0063 Th!s field alteration changed the main turbine steam admission logic from full arc (where all four turbine control valves throttle in parallel) to partial arc where only two valves open at low power, the third as power increases, and the fourth opens only at high power.
This change increases overall plant efficiency and shif ts the relationship between reactor power and first stage turbine pressure (first stage turbine pressure is higher for a given reactor power). Reactor pressure and the steam pressure upstream of the turbine control valves are the same for a given reactor power as before the change.
First stage turbine pressure is used as a mer.sure of reactor power for the rod-pattern controller.
To compensate for the increased first stage pressure, this supplement increases the low power setpoint from 124.5 psig to 138 psig and the lower allowable value from Idl.5 psig to 115 psig.
The new setpoint is based on calculations and data from post maintenance testing.
The plant technical specifications are being reworded to define the low power setpoint as a pressure setpoint using data from TCF008 testing rather than as a percent of rated thermal power.
Because the USAR describes the setpoint in terms of reactor power (20%), there is no change to the USAR description.
The new setpoints include conservatism to account for instrument accuracy, calibration uncertainties, and drift during the interval between calibration.
The setpoint change has also been reviewed to ensure that it does not result in spurious rod blocks during normal operating conditions.
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PAGE 9 CHANCE THE COACULAKT USED IN THE TREATNENT OF RAW VATER IN THE MAKEUP WATER SYSTEM Docurent Evaluated: FA WMFO?0 si Log utnber: 91 0060 This change modified the coagulant used in the treatment of raw water in the makeup watec system from sodium aluminate to a blend of polyaluminium chloride and cottonic polymers. The change is being made to improve the reliability of-water-treatment equipment since the replacement coagulant is less viscous and does not precipitate aluminum oxide which clogs lines, instruments, and pump:.
Water chemistry will not be adversely affected by this change.
The reactor grade water quality requirements currently in place will not be revised by this change.
REPLACE SERVICE WATER FLOW CONTROL VALVE Doctnent Evaluated: FA WSF008, $2 Log Ntsnber: 91 0016 The plant service water (WS) system provides cooling water to the plant
- chilled water (WO) system chillers.
Problems were experienced with the WS flow control valve (1WS079A) at the outlet of chiller OWOO2CA. The existing valve was not capable of withstanding the high 61fferential pressure experienced when service water was at low temperatures and low flow. The valve was being damaged by cavitation.
This change repinced the valve with a new type of valve better designed to control the pressure drop without cavitation damage.
This change also provided for a larger diameter instrument air line to the operator for this valve.
Since this operator only required an intermittent supply of air this change did not impact the instrument air system. The WO system is not required to achieve or maintain safe shutdown.
INSTALL PRESSURE TAPS IN PLANT SERVICE VATER SYSTEM TO VALIDATE FLOW BALANCE Doctnent Evaluated: FA Wsf009 & 51 Log NLeber: 91 0046, 92 0044 This change installed pressure taps at various points in the plant service water system.
The plant service water system provides cooling water required by station auxiliary equipment during normal station operation it is also the supply. to the shutdown service water system during normal station operation
- and shutdown.
The installation of pressure taps at selected locations in the system will allow-data to be taken for verifying system flow balance calculations.
The installation of these pressure taps does not impact the performance of the system.
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PAGE 10 INSTALL FLANGED PIPING IN SERVICE WATER PIPING TO FACILITATE CHILLER MAINTENANCE Docwent Evaluated: FA WSF013 tog mater: 91 0061 Flanged piping.was installed in the plant service water (VS) system to facilitate maintenance of the plant chilled water (PO)-system
'E' chiller.
The WO chillers are not safety related, also there are no safety design bases-for the VS system.
If WS becomes unavailable, the source of cooling water is-transferred to the shutdown service water system pumps automatically.
Failures of the WO chillers are not evaluated in the USAR, and this change does not introduce a new failure mode.
This change does rot affect the flooding analysis since Icakage from the spool piece would be no worse than that already considered for this area.
DECREASE SETPOINT OF MAIN CENERATOR HYDROGEN COOLERS RELIEF VALVES Docanent Evaluated: FA WSF014 tog Nwber: 91-0050 The main generator hydrogen coolers are cooled by plant service water (WS) and are provided with thermal overpressure protection relief valves.
This change decreases the relief valve setpoint from 170 psig to 130 psig.
The 170 psig was only achievable if plant service water was isolated while heat was being added to the cooler.
The maximum operating pressure of the cooler is 150 psig. Thus, the cooler could have experienced a pressure Sreater than its design pressure at the original setpoint. Also, tube sheet repairs resulted in reducing the design pressure to approximately 148 psig.
Failure of the hydrogen coolers is not evaluated in the USAR.
The failure of a hydrogen cooler would be enveloped by existing analyses in the USAR, REPLACE PLANT SERVICE WATER HEAT EXCHANGER FLOW REGULATING VALVES Docunent Evaluated: FA WSF015 tog Nmber: 92 0027 The plant service water (WS) system provides cooling wat.: to various systems and components in the plant.
The component cooling water (CC, system is an intermediate heat exchange loop between the WS system and systems which are potentially-radioactively contaminated, such as the fuel pool cooling and cleanup system. Valves 1WS018A and B regulate flow through the CC heat exchangers to maintain the CC at its desired temperature.
The valves have been experiencing cavitation producing vibration and noise, and c.cnstant breakdown of these valves. This modification replaces these valves with new aspirator valves. The aspirator-acts as a vacuum breaker which minimizes the cavitation inside the valve body.
Support modifications and other minor piping changes were included.
This change is expected to. improve the operation of these valves and does not adversely impact any equipment malfunction or accident analysis.
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PACE 11 REMOVE EDUCTORS AND ASSOCIATED SMALL BORE PIPING FROM THE SPENT RESIN TANK Docunent Evaluated: FA Wxt008 Log Nueer: 90-0054 This change removed eductors, spargers and associated piping from the spent resin tank.
This change was made because-the eductors, spargers and associated piping became plugged with resin during tank recirculation.
This change allows the spent resin tank to be recirculated without causing plugging of the recirculation line.
No failures of the spent resin tank are evaluated in the USAR.
No new type of malfunction or accident is created.
DELF.TE NONEXISTENT TEST CONNECTION FROM BREATHING AIR SYSTEM DRAVINGS Docunent Evaluated: FECN 27170 log Nueer: 91-0070 The breathing air (RA) system provides breathing quality air to stations throughout the plant.
It was identified that a test connection shown on the RA piping and instrumentation drawing (P&ID) had not been installed in the ficid. This design change deletes the test connection from the P&ID. The RA system performs no safety related function.
ALTERNATE COATING FOR DIESEL FUEL OIL STORACE TANKS Docunent Evaluated: Mod DG 062 Log Nuter: 92 0053 The diesel generator storage tanks and day tanks for Divisions 1, 2 and 3 are coated with a corrosion inhibitor. The coatings specified in USAR Section 9.5.4.2 are no longer available. The original paint manufacturer has specified an equivalent coating, which is compatible with the exising coating and will not react with diesel fuel oil or other fuel oil add 1*.ives.
The new coating is designed to handle aggressive solvents and oils.
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3 PACE.12 ALLOV REACTOR OPERATION THROUGH CYCLE 4 WITH EIGHT OF TWENTY-EIGHT SHROUD HEAD STUD ASSEMBLY BOLTS REMOVED Docunent tvaluated: Mod kB-028 Log Nuter - 92 0057 This change evaluated reactor operation through the end of cycle 4 with eight of twenty-eight shroud head stud assembly bolts removed and with three new bolts of a modified design installed. The shroud head stud assembly clamps the shroud head and steam separator assembly to the shroud.
The studs are disengaged-to allow for the removal of the separator assembly.
During installation of the separator assembly, the proper torque is applied to the stud remotely, using a bolt which extends from the stud to the top of the steam separators. Af ter the stud is torqued, the bolt is locked in place by a spring-loaded locking collar which has internal splices that mesh with external splices on the bolt near 'he head.
This prevents the bolt and stud from rotating during service.
The bolt has no other function during reactor operation.
During the second refueling outage, four of the twenty-eight shroud head stud assembly bolts were found to have sufficient wear of the locking splices to warrant their re aval since a potential loose part could re sult.
Continued operation through cycle 3 with these four bolts removed was justified in a safety evaluation reported in the 1991 annual report.
Due to continued wear of the 24 remaining bolts during cycle 3, it was necessary to remove seven additional bolts during RF.
In order to meet bolt spacing criteria assumed in the vendor's analysis three new bolts of a modifiec design were installed. The wear resultea irom feedwater flow impinging on the bolt shafts. Ultimately, this wear could render the locking collar assembly inoperable and significantly weaken the bolt shaft. Also, several other bolts and collars have slight wear of the locking splices but are still capable of performing the locking function.
Analysis showed that this condition (eight of twenty-eight shroud head stud assembly bolts removed and three new bolts of modified design installed) would not have an adverse effect on operation or response to accidents and transients.
Shroud head lift, stresses, and loose parts were considered.
ADDITION OF A FLUSHING CONNECTION TO THE SX-RH CROSS TIE Docuent Evaluated: Mod SX 032 Log Numer: 92-0 % 9 A cross tie between the shutdown service water (SX) system and the residual heat removal (RR) system is routed such that port sns of the line have trapped silt and restrict flow. The line is not used during normal conditions.
A flushing connection is requited to prevent silt build-up, The tee will be blind flanged during normal operation. The only time this cross tie is used is under post desi. i basis conditions should reactor pressure vessel flooding be needed. The addition of the flushing connection v' ensure its ability to
_ function.
PACE 13 ADDITION OF MANUAL ISOLATION VALVES TO THE DRAIN LINES FOR THE REHEATER DRAIN TANKS Docment Evaluated: Mod 10 007 Log Nunber: 92 0073 This modification added a manual isolation valve to the reheater drain tank turbine drain (TD) line.
This change provides double isolation which greatly improves the isolation capability required during normal plant operation and allows for increasect flexibility for the maintenance, The entire TD piping system is constructed to ANSI B31,1 Power Piping code and is categoris:cd as non seismic in the design specification.
Impacts on existing supports, piping and equipment as applicable were evaluated.
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CLINTON POWER STATION 10CFR50.59 REPORT-
-FOR TEMPORARY HODIFICATIONS FROM MAY 1991 THROUGH'MAY 1992
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PAGE.2 MAIN CONDENSER SCALI INHIBITOR CHEMICAL TREATMENT Docunent Evaluated: TM 91 007 tog Nuter: 91 0055 This temporary modification installed a chemical feed systern during the third refueling outage, This system feeds scale inhibitor to the main condenser circulating water system pump suction bay.
A hard calciur, carbonate scale was found in the main condenser tubes during a second refueling outage inspection, The scale impedes heat transfer and causes a taidsummer generating capacity loss of 6 MWe, Chemical treatment is required to prevent further scale buildup and generating capacity loss.
The-inhibitor being used is phosphoric acid.
The added weight of the feedpump skid was evaluated and determined not to affect the seismic qualification of the screenhouse which is Seismic Category I.
Interaction between the phosphoric acid and the sodium hypochlorite (currently used as a biocide) was evaluated and determined not to be a problem because of the short contact time in a once-through system.
Effects on plant materials were evaluated. Also, IP determined that there would be no detrimental ef fect on the environment due to use of this chemical.
OPTIMIZE TIMING O' REACTOR WATER CLEANUP SYSTEM FILTER PREC0ATING PROCESS Document Evaluated: TM 91-013 Log Nuter: 91-0053 This temporary modification changed the times for various steps in backwashing/precoating phases of the reactor water cleanup (RT) system filter /demineralizers.
The sequence and timing of the precoating process are not discussed in the USAR in detail; however, USAR 5.4.8.2 states that the time to take a filter off-line, backwas' and precoat is less than one hour.
This change allows the time to exceed one hour as required to optimize filter performance.
This is expected to enhance the performance of the RT filters by ensuring that the filter is evenly precoated over its entire length, by reducing the amcunt of resin _re3 aired to precoat one RT filter, and by enhancing capacity to absorb some ions (mainly chromates). The time required to take a filter off-line, backwash, and precoat is not part of the design
- bases, INSTALL PIPE SURFACE TEMPERATURE DETECTORS ON STEAM CYCLE SYSTEMS PIPING Cocument Evaluateo: TM 91-014 Log Number: 91 0113 Thirty-four steam cycle valves were instrumented to monitor heat dissipation which has the potential to significantly reduce plant electrical power output when leakage is experienced. This type of monitoring is referred to in the industry as a Potential Loss Monitoring Program (PLMP).
E1 ht of these valves 6
were instrumented via temporary modification 90-069 for the pilot PLMP during the second refueling outage.
The remaining twenty-six valves were instrumented during the-third refueling outage to expand the scope of the i
pilot' PLMP, for the purpose of further technical evaluation of this program.
Pipe surface temperature monitoring resistance temperature devices were installed by this temporary modification..The monitoring system is entirely nonintrusive in nature.
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PAGE 3 ISOLATE INSTRUMENT LINE TO PRESSURE SVITCllES VHICil CONTROL MOISTURE SEPARATOR /R~EHEATER VALVES-Docunent Evolueted: IM 91 015 tog Huntier: 91 0059 One of two pressure switches on an instrument line from the main turbine cross around stearn line failed, causing a steam leak in the turbine building.
A change was made to lif t a lead interrupting a logic signal from the two pressure switches in order to isolate the instrument line to the pressure switches.
As a result, a pressure transmitter on this instrument line was also isolated.
Cross-around steam pressure is used as an indication of power level. One pressure switch is set so that the moisture separator /rcheater (MSR) reheater drain tank emergency drain valves open below 20% power, The other pressure switch is set so that, below 10% power, the MSR moisturo separator drain tank emergency drain valves open; the main steam-to-MSR 1 solation valves close; and the MSR scavenging steam block valves close.
The -
2 MSR emergency drain valves fall open or automatically close at low power to purge air from the MSR during startup and to dry the MSR during turbine shutdowns.
This temporary modification required a lead to be lif ted to prevent t.hese automatic actuations while the pressure switches were isolated.
The pressure transmitter provides input for control of the MSR main steam control valves.
Ilowe ve r, since these valves were manually controlled, there impact as a result of isolating this transmitter.
was no INSTALL BYPASS PIPING AROUND INOPERABLE IS01ATION VALVE FOR OFFCAS liYDROCEN ANALYZER Docunent Evaluated: IM 91 016 Log watier: 91 0058 This temporary modification bypassed one of the main condenser offgas (00) system hydrogen (lig) analyzer isolation valves (N66 F085B).
The 112 analyzers measure lip concentration in the OG stream. There are two redundant analyzers in parallel.
Sample flow is provided by connections to the OG process piping and to the main condenser.
Condenser vacutun is used to draw the sample stream through the analyzers.
Each analyzer has two isolation valves on the vacuum side.
One isolation valve is on the analyzer itself and the other was added during construction.
In this circumstance, the "B" analyzer was operable but its isolation valve would not open.
This temporary modification installed piping from the "B" analyzer to a location between the two isolation valves of the "A" analyzer, This provided an effective flow path for the
".B" analyter, making it operable.
The isolation valve could not be repaired during plant operations since this could cause a loss of condenser vacuum, resulting in a plant trip.
The temporary piping is constructed to the same codes and quality the original piping, as
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PAGE 4-DISCONNECT INTERFACE VIRING BETWEEN LIQUID RADWASTE DISCHARGE RADIATION MONITOR AND PANEL IN RADWASTE OPERATIONS CENTER Doctanent Evaluated: TM 91 017 Log Ner 91 0066 The liquid radwaste discharge monitoring instrument (OPR04D) measures the radiation concentration of the liquid radwaste discharge into the plant service water discharge header.
Its main function is to provide a signal on high radiation level to close the liruit radwaste discharge isolation valve.
Various controls and annunciators are provi<ied in the main control room, the radiation protection office and the radwaste operations center and are described in USAR Sections 11.5.2.2.6 and 7.7.1.19.4.
This temporary modification disconnected the wiring between the monitor and the panel in the radwaste operations center to eliminato electrical noise that has caused the monitor to be declared inoperable.
Sufficient controls and annunciations remain in the main control room and radiation protection office to provide for operation of liquid radwaste discharges.
The capabili*; for the liquid radwaste monitor to initiate closure of the discharge isolation valve upon high radiation alarm was retained, t
INSTALLATION OF PRESSURE CAUCEE, AT THE CIRCULATING UATER PUMP DISCHARGE Doctsnent EvaluatN: 1p 91-028 Log Ntsnber: 91 0087 Pressure gauges were installed on existing capped root valve lines on the discharge piping of each circulating water system pump (three pumps).
This was done to provide pressure data on the circulating water system.
The data will be used to calculate condenser thermal performance.
Since USAR Figure 10.4-3 shows the lines for these valves as capped (ICU008A, B, and C), this is a change to the facility as described in the USAR.
It was judged that the temporary modification would not increase the probability of a loss of-condenser vacuum since failure of the modification would not cause an appreciable loss of circulating water system flow.
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INSTALL NEW SODIUM HYPOCHLORITE INJECTION SYSTEM Doewent Evaluated: IM 91-031 Log Ntater: 91 0093 Temporary sodium hypochlorite injection equipment for the circulating water and service water systems was originally installed under temporary modification 88-044.
A new sodium hypochlorite storage tank and associated isolation valves and piping were installed to replace the tanks and valves installed in the original temporary modification because they had developed.
Icaks. Although this change does not affect the facility as described in the USAR, a 10CFR50.59 safety evaluation was performed because of potential interaction with safety-related equipment and tructures in the screenhouse.
The sodium hypochlorite addition reduces biofouling and the intrusion of corbicula into the circulating water and service water systems.
The new installation was analyzed for the effects of spills and seismic and wind loadings on the screenhouse slab.
Failure of this temporary modification will not impact the fire protection or the shutdown service water system because of separation and the provision of a berm to contain spills.
PROVIDE COOLINb WATER TO THE "A" FIRE PUMP DIESEL DURING POST-MAINTENANCE TESTING Doctsnent Evaluated: TM 91 032 Log Ntster: 91 0094 The "A" firepump is the normal source of cooling water for the diesel driven fire pump.
During post-maintenance teccing of the diesel engine, the normal cooling water supply was disconnected and it was necessary to provide a supply of cooling water from an alternate source. This temporary modification connected a cooling water supply to the diesel engine from a drain valve on the service water system. Hoses were routed through the nonsafety-related areas of the screenhouse and connected to the pressure regulating valve on the supply side of the diesel engine heat exchanger.
The exchanger requires approximately 30 CPM of cooling water flow.
The service water system flow was not significantly affected.
During this maintenance effort, the "B" fire pump was operable, and pump "C" was available as a back-up.
An operator was stationed in the vicinity of the open service water system drain valve in the event of a hose break t
BLOCK CLOSED RELIEF VALVE TO PROTECT PERSONNEL WORKING IN THE CONDENSER Docunent Evaluated: IM 91-036 Log Nmber: 91 0116 This modification blocked closed relief valve 1B21-F408 located on the main
- steam (MS) supply line to the auxiliary steam (AS) system, which' discharges to the main condenser.
This was done to protect personnel working in the main-condenser during an outage.
The upstream MS valves were closed so that this portion of the system was not exposed to MS pressure.
Pressure surges originating from the AS reboilers would be relieved by the relief valves on the reboilers.
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PAGE 6 OXYGEN INJECTION SYSTEM Docment Evaluated: TM 91-037 Log Nwter: 91-0106 A system for injecting oxygen in the condensate pumps suction piping was installed.
This was done to minimize corrosion of the carbon. steel and low-alloy steel surfaces in the condensate and feedwater systems.
Studies have_
indicated that oxygen concentrations between 20-50 pg/L allow a stable oxide layer to form on the piping surfaces.
Prior to this modification, feedwater oxygen level was typically 18 g/L.
This system was designed to bring that level up to 40 pg/L.
During the time the injection system is in operation, data on pipe wall thinning and crud formation will be taken.
Based on'the industry data, oxygen injection will minimize the formation of new corrosion products and retard corrosion product release to feedwater and subsequently to the reactor, The installation was evaluated for effects of missiles and fires, and resulted in no new potent *.l equipment failures important to safety.
TEMPORARY REPLACEMENT OF TVO REACTOR VATER CLEANUP SYSTEM PIPE SUPPORTS Document Evaluated: TM 91-040 Log Nwter: 91 0122 This temporary modification converted two reactor water cleanup (RT) system pipe supports from constant supports to ri id supports.
This was done to 6
support installation of temporary shielding on the RT piping in the drywell during the third refueling outage. The temporary shielding was installed only during Mode 4 (cold shutdown) and Mode 5 (refueling) and was removed prior to the end of the outage. An evaluation determined that all pipe stresses were within ASME Section III allowable values and that all pipe supports and auxiliary and structural steel were structurally adequate.
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DRAIN THE DIVISION II DIESEL GENERATOR FUEL OIL STORAGE TANK-Document Evaluated: TM v1-045 Log Number: 91 0131 l
During the third refueling outage, the Division II diesel generator (DG) fuel i
oil storage tank was temporarily modified to drain the_. tank.
Also, an air mover was attached to the normal flame arrester connection.
These changes-were done while the DG was out of service for maintenance.
Since the temporary modification was installed while the DG was out of service for maintenance; the fuel oil system was restcred to design configuration; and the fuel oil was sampled and analyzed prior to restoring the tagout, an unreviewed-safety question was not involved.
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PAGE 7 INSTALL MONITORING EQUIPMENT FOR TROUBLESHOOTING SERVICE AIR COMPRESSOR Docunent tvstuated: TM 91 050 Log kster 91 0111
-This temporary modification installed four pressure transmitters and one monitor to a service air compressor. The service air system has three compressors that supply service air and instrument air for maintenance, instrumentation and controls throughout the plant. This temporary modification installed tees or connections at different tubing locations on one compressor, to allow data collection to help determine why this ccipressor has had pressure surge problems. All installed devices were rated for system pressure, voltage, und amperage.
The operation of the temporary modification monitoring system did not impact the compressor operat. ion.
The two remaining service air compre. ors were not affected, nor was there an increase in the probability of occurrence or consequence of a loss of instrument air.
PLANT SERVICE WATER SUPPLY FOR CONDENSER TUBE CLEANING EQUIPMENT Docunent EvaluaW: TM Y1 051, R1 Log heters v2-0012, R1 This temporary modification installed a temporary water supply from the plant service water (VS) system to the condenser tube cleaning pumps used during the third refueling outage.
These pumps shoot scrapers through the main condenser tubes to clean out all soft and hard deposits. Temporary hoses were attached to seven drain valves in the WS system.
Revisior 1 of the temporary modification installed two additional hoses attached to WS drain valves.
These hoses were removed when cleaning was complete, and the configuration was restored to its normal ctate. The condenser tube cleaning equipment uses approximately 150 CPM of water when operating which is negligible compared to the total flow to the affected components; therefore, none of the operating characteristics of these components were affected.
In the event of a shutdown service water (SX) system automatic start, all but one drain valve would have been isolated from the SX system.
This event would not have affected the operation of the drywell chillers, and would not have depleted the Ultimate Heat Sink water inventory, l
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PAGE 8 DISABLE ROD CONTROL AND INFORMATION SYSTEM ROD BLOCK Docunent Evolunted: TM 91 052 Log W M wra 91 0112 The intermediate range monitors (IRM) are neutron detectors used to monitor,.
record, and provide sipials to both the recctor protection system (RPS) and
-the rod control and information system (RC&IS) during reactor startup, heatup, and shutdown. The IRM channels produce a trip of both the RPS and RCIS rod block when placed in the inoperable position, To meet the requirements of the technical specifications, the channel must be pieced in the tripped position creating a scram signal for that channel. This charges-the scram logic to one-out-of-three since one trip signal is already present, However, since the rod block occurs when any one channel is in the tripped condition, there is a continuous rod block present.
This temporary modification placed'the IRM r channel in bypass for the rod block signal while not affecting the scram signal.
The bypassing of the a single IRM channel reduces the associated channel trip logic to one-out-of-one.
However, the bypassir,6 of the IRM C red block signal is allowed within technical specifications.
The reason for the temporary modification is that IRM C is reading erratically (0 to 125%) and is generating both reactor scram and rod block signals.
Ts allow the plant to startup while satisfying the technical specifications, requires only six operable IRMs; therefore, IRM C rod block is being bypassed.
The continuous rod withdrawal evaluation in USAR 15.4.1.2 is not considered credible during startup. USAR Chapter 7 lists the rod block as f art of the power generation design bases and not the safety design bases No other IRM channel can be tested with the temporary modification installed since testing would satis fy the two-out-of-four trip requirements causing a scram, The evaluation showed that failure of the temporary modification due to a seismic event would generate e rod block.
REMOVE THE CURRENT FLOW COMPARATOR INPUT TO THE PROLONCED IDSS OF STATOR COOLING TURBINE GENEkATOR RUNBACK CIRCUIT Docunemt Evaluated: TH 91-064 Log Number: 91 0128 This temporary modification removed one of the three inputs to the prolonged loss of stator cooling turbine generator runback circuit. This circuit protects the generator from damage due to loss of cooling water to the generator stator.
The prolonged loss of stator cooling runback is described in the USAR, although the inputs to this circuit are not.
Several spurious turbine generator runbacks occurred due to failure of the current flow comparator, With the remaining inputs to the runback circuit capable of performing their function, the turbina generator will still trip.if the runback is not completed within a set tiee.
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PAGE 9 FABRICATE MOUNTING BRACKET AND WIRING HAKNESS FOR ROD POSITION INFORMATION
. SYSTEM POWER SUPPLY MODULE
- Doc'aent Evaluated
- TM 91-067 Log Water 91 0130 Th!s temporary modificatior. replaces a mounting bracket and wiring harness for a failed power supply ir rod control and information system position indication' circuit.
...o mount ng of the temporary power supply module is i
identical to the-original and will not become a missile during a seismic event. Wiring sizes of the new harness were also evaluated as acceptable.
B14CK '?EN CONTAINMENT IS01ATION VALVES FOR SERVICE AI't, INSTRUMENT AIR, i
BREATHING AIR, AND EQUIPMENT / FLOOR DRAINS Document Evatusted: TM 91 069 Log Nuter: 91 0121 This temporary modification blocked open several air-operated containment isolation valves using stem col'ars so service air, instrument air, breathing air, and eculpment and floor di
.a flow could be provided during the third refueling outage.
The air supply to these valves was removed while the instrument air system was, out of service for maintenance. The stem collars were installed to prevent the valves from closing due to the loss of air.
Also, jumpers were installed for the equipment and floor drain sump pump circuitry to prevent the sump pumps from tripping.
These containment isolation valves were not required to be operable in Mode 4 (cold shutdown) or Mode 5 (refueling and core alterations) during the period when this temporary modification was instelled.
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PAGE 10 DEFEAT FUgL llANDLING I LATFORM INTERLOCKS Docment tvaluatedt tM 91 070 tog meter: 91 0126 The norrnal operating mode of the fuel handling platforra is for inoving new and spent fuel between the inclined fuel transfer systeru (1FTS) cart and the spent fuel pool. These transfers are completed under water.
Interlocks are provided to prevent refueling bridga crane movement through the gate between the spent fuel pool and the IFTS transfer pool if the refueling inast is not centered on the gate.
Other interlocks prevent the IFTS cart movernent into the IITS transfer pool area unless the upender is in tho vertical position.
This temporary modification set up the fuel handling platform to roove new fuel from the storage vault to the spent fuel pool.
For this operation, the fuel handling mast is stowed in a horizontal position, and the auxiliary !.oist, rather than the mast, is used to transport new fuel.
The new fuel is lifted vertically until the bottom of the fuel bundle is above the top of the fuel pool.
The fuel is then moved laterally over the IFTS transfer pool area and then lowered to its tempolcry location in the spent fuel pool. This temporary modification defeated the interlocks in preparation for the new fuel transler.
Also operation from the auxiliary hoist / monorail pendant would normally re,uire a second operat.or to pres i the bridge override pushbutton at the snain control panel.
A jumper was installed to allow operatinC the bridge from the auxiliary monorail pendant without. the need for the second ope,ator.
During this operation, the new fuel was not moved over spent fuel.
3 LOVER SETPOINT FOR FUEL POOL COOLING AND CLEANUP PUMP MOTORS COOLING WATER LOV FLOW TRIP Docment Evolueted: IM 92 002 Log meter: 92 0028 During the 4160 volt 1A bus outage during the thiro refueling outage (RF-3),
the service water (SX) system and component cooling water (CC) system were removed from service at the mme time.
During thin time, shutdown service water, which is colder than ) system water, was used to cool the fuel pool cooling cud cleanup (FC) system pump rnotors.
Technical specifications require that reactor vessel we..
In the upper containment pool he kept above 68'F t o maintain the reactivity shutdown snargin. The US.'J1 requires that spent fuel pool water r
.o be kept above 68'F for the s ee reason.
Furthermore, because RF-3 was conducted during the spring, the lake water (shutdown service water) 3 was colder than normal.
Because of t..a colder temperature of the SX water, it was necessary to educe SX flow to the FC heat exchangers to prevent cooling the pools below 6-'F.
Pool temperatures and SX flows wert monitored and v
controlled adminis.trative.ly. This temporary inodification specifically lowered the trip set points on the FC pump rnotor's cooling water flow indicatirg switches from 25 GPM to a CPM ir, order to prevert spurious trips of the pumps with the reduced SX flow.
Calculations demonstrated that 8 GPM is adequate to remove the heat load with the cooling water at or below S0*F, which was the
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case while this temporary roodif t :ation was installed.
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PAGE 11 TEMPORARY ELECTRICAL BACKFEED FROM SWITCllYARD 'IHROUGli Tile MAIN POWER TRANSFORMERS TO Tile UNIT AUXILI ARY TRANSFORMERS Docmient tweluated: IM 92-004 tog W eber: 92 0005 t
Due to a failure of the "B" phase main power transformer (MPT), Jt was necessary to backfeed electrical power frorn the switchyard through the MPTs to the unit auxiliary transforners to test the re placemer,t MPT.
This ternporary rnodification removed the reverse power relays, adjusted the power irnbalance relays to give maxirnurn protection, installed a jumper to keep the turbine running on its turning gear, and installed a jumper to keep the turbine 4
control valves closed.
The generator was off line and physically disconnected frorn the isophase bus ducts during this installation.
BIACK OPEN OFFCAS DESICCANT DRYER INLET VALVE Docwent f valuated: im 92 011 tog hatter: 92 0017 The off gas systern has two desiccant dryers in parallel. The dryers dry the off-gas offluent to approximately -50*F dew point prior to going to the gas cooler and charcoal beds.
Moisture is removed froin the of f-gas to prevent freer.ing and blocking the charcoal beds. One desiccant dryer is in service while tho other dryer is being regenerated or is in standby.
During a normal transfer between the inservice and standby dryers, the "B"
dryer inlet valve, 1N66 F012B, failed to open.
This temporary modification blocked this valve open.
Operators used the manual inlet isolation valve IN66 F121B to manually isolate /open the dryer. These valves are shown on USAR Figure 3.6 1, Sheet 75.
Administrative controls were implemented to ensure proper operation of the IN66F121B valve.
RELOCATE llYDR0 LASERS USED FOR LEAK RATE TESTING DURING RF-3 Dotwent tvaluated: 1M 9/ 012 tog katoer: 92 0023 Two high-pressure hydrolasers are used during refueling outages to conduct is;scrvice inspection leak rate tests.
This temporary rnodification relocated the hydrolasers to the northwest area of the fuel building, elevation 755',
during RF 3.
This installation was reviewed for its effect on the Fire Protection Evaluation.7eport (FPER), for possibio effects:of water spray due to-hose failures, for internal flooding, and for structural and seismic considerations.
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j PAGE 12 BLOCK OPEN OFF CAS DES 1CCANT DRYER INLET VALVE Docawnt (volunted: 1M 92 013 top wmter: 92 0019 l
During a normal transfer between the inservice and standby dryors, the inlet valve of t he "A" dryer failed to fully open.
This temporary modification blocked valve IN66 F012A open. A second manually-operated valve was used to isolate the "A" dryer during its regeneration cycle. A similar problein occurred on the "B" dryer inlet valve four days. prior to t his occurrence.
Administrative controls were placed on operation of the manual inlet isolation valve.
Failure of operatiotu. to properly manipulate the valve could cause a loss of off-gas flow, and subsequently, a loss of condenser vacuurn which is an accident evaluated in the USAR (Section 15.2.5.).
A loss of vacuum caused by operator error in manipulating this valve would be less severe than that evaluated in USAR 15.2,5 because the vacuum decay would be slower.
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INSTALL WINDOW ON DRYVELL AIR PARTICULATE MONITOR Dotwent Evaluated: TM 92 018 tog kmder 92 0022 The drywell air particulate monitor provides a secondary method for detecting a leak in the rer.ctor coolant. pressure boundary within the drywell.
The air t
particulate sample panel filtert ;ut representative sampics of gaseous and effluent particles, trap,)ing the particles on filter papet.
A gear drive assembly advances the paper, moving the sample to a beta scintillation detector. As originally installed, the paper path was totally enclosed and not visible without opening up the pane?,
On several occasions the paper advance mechanism failed.
Because the failures could not be observed, they were not discovered until the nett equipment disassembly, which was typically the preventative maintenance task to replace the filter paper.
The equipment was therefore inoperabic without the knowledge of the operators and in violation of the technical specifications.
This temporary modification installed a clear viewing window in the cover of the compartmenc for the supply filter paper reel. This change allows direct observation of the reel which can only_ rotate if the paper is moving.
This change allows identification of a failure in time to effect a repair and avoid a violation of technical specifications.
The temporary modification was evaluated for seismic and environmental considerations and found acceptable.
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PAGE 13 MOBILE TOOL DECONTAMINATION FACILITY Docunent tvatusted: im 92 023 tog Nuntaer 92 0030 This temporary modification consisted of a mobile tool decontamination facility during the third refueling outage. The facility was self-contained and was composed of two 40-foot by 8-foot vans and one 40 foot flatbed trailer.
The system cleaned radioactively contaminated tools by means of a high velocity delivery of carbon dioxide (CO ) pellets.
The vans contained 2
the pellet cleanin6 enclosures, ventilation equipment, pelletizer, nir drying equipment, electrical distribution equipment, and a clean storage area.
The flat bed trailer held a 14-ton CO2 storage tank.
The facility was set up outside the power block at the southeast corner of the fuel building.
The safety evaluation addressed:
the possibility of airborne releases of gases and particulates; the potential risks to main control _ room habitability and diesel generator operability due to rupture of the CO2 tank or failure of components such as the safety relief valve; effects on secondary containment; additional loads on plant structures; solumic events; tornados; and fire protection.
INSTALL REACTOR VATER LEVEL TRANSMITTER SINULATOR ON ANALOG TRIP MODULE Docment tvaluated: 1M 92-032 Log htenber 92 0047 During the third refueling outage (RF-3), due to maintenance activities, it was necessary to remove the sensing line to reactor water level transmitter 1821-N080D.
This temporary modification disconnected the sensing line and installed a transmitter simulator on the analog trip modules (ATM) 1B21-N680D and IB21-N683D.
USAR Figure 5.1-3 shows the connection between the transmitter and the ATH.
This ATM provides actuation signals to the reactor protection system (RPS) and the containment and reactor vessel isolation control system (CRVICS).
This instrumentation is covered by Technical Specifications 3.3.1 1 4, 3.3.1 1.5, and 3.3.2-1.5.C; however, the temporary modification was installed only during modes 4 or 5 when this instrumentation is not required to be operable.
Normally the logic for this trip is two out-o f - four. Without installing the transmitter simulator, this channel would be in a tripped condition meaning that the logic would be one out-of-three.
This i
uould make the trip more susceptible to inadvertent isolations of RHR shutdown tooling.
By installin6 the transmitter simulator and inserting a simulated 4gnal, the logic was effectively two-out-of three.
This change inhibited automatic isolation of several residual heat removal (RnR) outboard containment isolation valves including some in the shutdown cooling loop.
Manual isolation was still available using an isolation pushbutton or valve.
handswitches.
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i PAGE 14 REVISE SEQUENCE OF VENTING Tile REACTOR WATER CLEANUP SYSTEM FILTER /DEMINERALIZERS i
Docuent Evoluete<t: in 92 034 tog s w er 92 0054 The reactor water cleanup system filter /demineralizers remove solid and dissolved impurities from reactor coolant.
The filter /demineralizer units are pressure precoat type filters using a filter media and mixed ion exchange resins.
Problems were experienced with the sequence of venting the units prior to precoating.
Previously the vent valve closed first, allowing entrapped air to become pressurized in the filter /demineralizer dome area.
Subsequently, when the precoat return valve opened, the precoat tank would overflow creating a contamination area at the 803-foot elevation of the contairment building. The problem was controlled administrative 1y by stationing an operator to manually throttle the precoat return valve.
This temporary modification corrected the problem by revising the length of various timers so that the filter /demineralizers vent valve closes last.
The timing sequence is given in USAR Figure 5.4 19, Table 11.
Another change disconnected the precoat tank dust collector, which was originally installed to remove nuisance dust particulates from a previous precoat material which is no longer used.
DRAIN Tile DIVISION III DIESEL-CENERATOR FUEL OIL STORACE TANK Docuwnt Evaluated: 1M 92 035 Log Nder: 92 0051
.A temporary modification was installed which provided a method for draining the Division III diesel generator (DG) fuel oil storage tank so that the tank could be cleaned during the third refueling outage.
Also installed was-an air mover attached to the normal flame arrester connection.
These changes were made while the DC was removed from service for maintenance. An unreviewed safety question was not involved since the temporary modification was e
installed only while the DC was removed from service for maintenance, the fuel oli system war restored to design configuration, and the fuel oil was sampled and analyzed prior to restoring the DG to an operable status.
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CONTINGENCY CROSSTIE OF DC ELECTRICAL BUSES DURING DIVISION 1 BATTERY REPLACEMENT
-_d Document tvaluated: TM 92 037 & R1 Log Weter: 92 0050 & R1 Modification DCF004 replaced the Division 1 125 volt DC battery with a new, larger battery during RF 3.
While the battery was being replaced, the Division 1 DC loads were powered by the battery charger. A failure of the battery charger at this time would result in a power loss on the 1A bus.
This would result in power being lost to the normal supply of the Division 1 nuclear system protection system inverter, several reactor core isolation cooling system valves, and the 125 volt DC 1A distribution panel.
This temporary modification was installed as a contingency plan in the event of a battery charger failure.
It consisted of a crosstie between the 1A and IF buses so that the IF battery charger could be used as a secondary source of power for the 1A bus.
The connection complies with the requirements for circuit protection for IE buses connected to non 1E sources.
Administrative controls were in place to maintain loads on the IF bus within the bus design capability.
This change did not involve a change to the technical specifications.
It was in alled while the plant was in mode 4 or 5 when only one division was required to be operable.
A revision to TM 92 037 allowed the operational (non contingency) use of the temporary modification. The temporary modification was utilized to ensure that a primary and secondary source of power were present at the Division 1 bus.
This was done by connecting the Division 1 battery charger in parallel with the non-divisional 1F battery and hattery charger upon installation of the crosstic, and then turning off the Division 1 battery charger. This ensured that power was maintained at the Division 1 bus.
The parallel connection of these power sources was used for the limited time required for switching the circuit breakers.
The non-divisional 1F bus has 17,200 amps of avellable fault current.
With the Division 1 charger also connected to the Division 1 bus during this crosstle, an additional 375 amps of fault current is introduced for a total of 17,575 amps.
This value of fault current remains within the ratings of the breakers on the Division 1 bus, except for the breaker located in compartment 13A.
This spare breaker was not used during this temporrry modification installation since all loads on the Division 1 bus were shed with the exception of the 125-volt DC distribution panel and the emergency lighting cabinet. The expected load for this crosstic includes running the Division 1 diesel generator.
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PAGE 16 DRYVELL TEMPERATURE MONITORING (Elevation 800-Foot Azimuth 22-Degrees) TO DETERMINE LOCALIZED 110T SPOTS.
Docuwnt tvolunted: 1M 92 040 tog kwters 92 0071 Condition Report (CR) 1-92-03 016 identified damage to an instrumentation cable (1R116C) for a reactor core isolation cooling systern check valve. Also, CR 1 89 01-016 and 1 90 10 070 identified damage to cables 1R116C and 1RI16J during the firr,t and second refueling outages. This temporary roodification implemented the action plan to inonitor temperature profiles in the drywell (elevation 800-foot, azimuth 22-degrees).
This teroperature monitoring system included cables and resistance temperature desices.
Hardware will be mounted on the junction boxes and conduits.
The temperature will be monitored from the containment building throu6h permanent cables routed through an exiting drywell penetration. This change has no istpact on any permanently installed temperature elements.
O INSTALLATION OF BLIND FLANGE ON CONTROL ROOH llEATING VENTILATION AND AIR CONDITIONING (HVAC) CilARCOAL BED DELUCE LINE Docuwnt tvaluated: IM 92 043 tog Nmter: 92 0060 This temporary inodification installed a blind flange on the OVC075B filter unit charcoal bed deluge line to make the control room llVAC train "B" operabic.
This change allowed maintenance to be perforrued on valve 1SX076, assuring that the control room llVAC "B" train remained operable.
A blind flange on the OVC075B charcoal bed deluge connection irupaired automatic deluge fire protection capability, but compensatory measures were in place per CPS procedures, llourly fire watches of this area were performed.
The blind flange ensured Icak tightness of the control room ventilation system filter unit.
BLOCK OPEN VALVES IIA 005, 11A008, 1SA029, AND ISA032 Docuent Evaluated: 1M 92 050 tog kaitier 92 M76 Thin modification blocked open drywell isolation valves 11A005,11 A008,1SA029 and 1SA032, on the service air and instrument air systems during the third refueling outage. This was done to perform maintenance on t.he limit switches.
These valves fail closed when cont.rol power is lost.
These valves were blocked open with a stem collar.
By mechanically blocking opeu these valves, the associated systems could still be utilized. This chango disma d ;ne automatic isolation of the valves as mentioned above, however, these valves were not required to be operabic during this mode of operation.
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PAGE 17 BLOCK CLOSE VALVE DURING NITROCEN PURCE OF CilARCOAL ADSORBERS i
Docment Evaluated: IN 92-054 Log WW4r: 92 0082 i
This temporary modification blocked closed inlet valve IN66F131 to the refrigerated absorber vault in the offgas system. The purpose for maintaining i
the inlet valve closed was to allow purging of the charcoal adsorbers with nitrogen while eliminating the potential for inadvertent. introduction of oxygen into the charcoal adsorbers until the purging operation was completed.
The valve was restored to its normal state once the purging operation was completed. The change allowed controlled purging of the adsorbers. The system was operated in the bypass mode as described in USAR Section 11.3.2.6.3.
The nitrogen purge was maintained per CPS Procedure 3215.01. The nitrogen flow prevented backflow of offgas into the charcoal adsorbers.
Overall, system operation remained per design as described in the USAR Section 11.3.2.1.6.3..
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4 CLINTON POWER STATION 10CFR$0.59 REPORT FOR PROCEDURES AND DOCUMENTS FROM MAY 1991 TilROUCll MAY 1992 I
PAGE 2 IlELIUM LEAK TESTING Dotwent Eyaiunteds tPS Procedure 2B00.11 24, 2800,11D003 a0 tog Nmner: 92 0080 This procedure revision allows the use of Sulfur 11exafluoride (SF )
("
6 identify condenser tube leaks with the main condenser at 4 circulating water system in service.
For any single injection, SF6 will be injected into the circulating water system at a rate not to exceed 15 standard cubic feet por minute for one and one half minutes.
If condenser tube leakage exists, the SF6 gas will be detected in the offgas system strearn through the use of an SF6 gas detector.
Il condenser tube leakage is detected, one condenser waterbox will be isolated and drained at which time helium will be utilized as the tracer gan to identify which tubes are Icaking.
Condenser tube leak detection, using SF, has been utiltred at numerous othet nuclear facilities.
6 The nuclear steam supply system vendor evaluated the use of SF6 for condenser leak testing at CPS and recommended the method described above, Chemistry sampling of reactor coolant will monitor conductivity, chloride and sulfate levels to ensure these parameters reinain within the appropriate levels speeliicd in CPS procedures.
RAISE MAIN TURBINE BYPASS LOAD LIMIT FROM 105% TO 110%
Docunent Evaluated CPS Procedure 3004.01 R12, PDR 91 C766 Log We ber: 91 0062 During normal plant operation, steam pressure is controlled by the turbine control valves which are positioned in response 'co the pressure regulator signal. The turbine control valve dernand signal is limited to that value required to fully open the turbine control valves. Thus, if the pressure control system requests that. additional steam flow be relieved from the reactor when the control valvos reach wide open, a control signal will cause the main turbine bypass valves to cpen.
The limit for the pressure setpoint is called the load limit. This procedure revisien raises the load limit setpoint frcm 105% to 110%,
ducing periods when the moisture separator /rcheaters are not available.
At 100% reactor power, with the moisture separator /rcheaters not in service, more steam is sent to the high pressure turbine. To prevent the main turbine bypass valves from opening with the turbine control valves fully open, the load limit. was raised.
This change allows increased steam flow to the main turbine within its design capability as described in USAR Section 10.2.
Turbine control valve close stroke time is not affected by this change.
The main turbine will still be operated below the flows and pressures used in the accident analyses.
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PACE 3 OPERATE FEEDWATER PUMP MINIMUM FLOW CONTROLLER IN MANUAL ONLY Docteent tveluatedi CPS Proce&re 3101.01 R9 tog Wteder: 91 0072 USAR Section 10.4.7.5 states that the reactor feedwater pumps are provided with flow controls that automatically regulate pump recirculation.
The pumps require a minimwn flow of $000 CPM cach.
To provide minimum flow during periods of low feedwater flow, a miniinurn flow recirculation line with discharge to the main condenser is provided.
Controls are ptovided which allow operation in either automatic or manual.
!!ccause of problems encountered with the automatic mode, operations has opted to operate the systein with the minimurn flow controller always in manual.
This change deleten the USAR descript. ion of the l'W purnp minimum flow control as auto natic.
This USAR change reflects - che ge to the operating procedure that. states autornatic operation is not allowed.
None of the USAR Chapter 15 accident analyses are affected by this change.
PROVIDE INSTRUCTIONS FOR SUPPLYING CYCLED CONDENSATE TO CONTROL ROD DRIVES DURING PIANT OUTAGES Docunent ivaluated: CPS Procedure 3304.01 A14 tog Wtshr: 91 01?5 This procedure change allows the use of cycled condensate system water to provide cooling water to the control rod drive (CRD) inochanisms.
Continuous flow of high quality water is desired to cool the mechanical seals.
- Normally, the CRD pump oil coolers are cooled by the turbine building closed cooling water system which is cooled by the plant service water (WS) system.
During plant refuellog outagen, when it is necessary to shut down the WS systern to perforra maintenance, the CRD pumps must be shut down since their oil toolers will not be supplied with cooling water.
This procedure change allows high quality water to flow through the CRD mechanisms during WS system outagen through a cross connection to the cycled condensato systern.
Flexible hoses wfth in line filters are used to raake the cross connection.
If a hose were rut or were to burst, installed check valvoa would prevent backflow from the reactor vessel.
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,e PACE 4 USE OF LOV PRESSURE COOIANT INJECTION (LPCI) FLOWPATH FOR SHUTDOVN COOLING Doomnt Evolusted: cps Prxedure 3312.01 R17 Log Erber: 92 0043 i
The residual heat removal (RHR) system is designed to remove decay heat from the reactor during shutdown conditions in the shutdown cooling (SDC) mode.
Either R11R loop can be used depending on the decay heat load and the required l
cool down rate.
In the normal lineup of_an SDC Inop, suction is taken from j
the reactor recirculation (RR)
"B" ptump inlet line, pumped through the respective RilR heat exchanger (HX), and returned to the reactor vessel through one of the two feedwater (IV) lines.
CPS Procedure 3312.01 was revised to allow t.he use of the 1.PCI injection lines for SDC return to the reactor vessel during the second refueling outage (RF+2).
The configuration would allow one RHR loop to be operating in its normal SDC lineup with return to the reactor vessel through one of the tv lines.
The other loop would be lined up as a backup shutdown cooling system with return through its respective LPCI injection line.
This would enabic the second IN lire to be shutdown for maintenance.
'Ihc safety evaluation doctunented t.he availability of the LPCI flovpath during future refueling outages.
In late 1981 or early 1982, the Kuo Sheng (BWR 6) plant experienced damage to an in core instrument tube and surrounding fuel bundles due to jet impingement on the core while using LPCI injection piping for SDC. A change was made during CPS construction to add flow deflectors at the injection nozzle, to replace the in core instrument tube with a strengthened tube and to instruct operators to use LPCI only for accident or emergency situations. Allowing operation of SDC through an LPCI injection line in which suction la taken from the suppression pool requires several restrictions to be placed on using the LPCI lines for SDC. The time allowed for operations in this mode was limited to less than 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
In additfor, inspection of the six fuel assemblies in the immediate vicinity of the injection nozzle would be required prior to startup if this lineup were used. Also, flow rates were restricted holow 6000 GPM.
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4 PAGE 5 PROVIDE FOR ALTERNATE SilUTDOWN COOLING HETil0D USING UPPER CONTAINHENT POOL HAKEUP DUNP LINES Doctanent Evaluated: tPt Procedure 3312.02, a3 tog tuntser: 92 0031 This procedure revision adds another method for alternate shutdown cooling (ASDC).
This method, which is used during refueling outages with the reactor pressure vessel (RPV) head removed, provides for returning water froin the RPV through the upper containment pool dump lines to the suppression pool.
The previous ASDC methods enabled water to be returned to the suppression pool using three of the main steam relief valvo discharge lines. Any one of t.he five emergency core cooling system pumps inay be used to take suction froin the suppression pool and inject into the RPV.
Several prerequisites assure safe operation in the new alternate modo.
No fuel assemblies are allowed to be stored in the upper pools.
No fuel semblies or control rods are allowed to be handled within the RPV, All af ueling operations in t.he containment building are suspended.
The nuppression pool water level is lowered to accommodate the additional volume of water.
If the high pressure core spray pump is used, it is aligned to the suppression pool to prevent addition of water from the reactor core isolation cooling storage tank to prevent overflow of the drywell welr wall.
Sinco fuel movement is not allowed while this flow path is in use, there is no impact on the analysis of e dropped fuel bundle in the containment or any other USAR analysis.
ALTERNATE SUPPRESSION POOL COOLING USING FUEL POOL llEAT EXCllANGERS Docunent [ valuated: tPS Procedure 3318.02, R1 tog Nuntiert 92 0014 This procedure provides an additional method of decay heat removal from the reactor vessel.
This procedure provides for an abnormal supprossion pool cooling configuration in which suppression pool water is pumped through one of the two fuel pool cooling (FC) heat exchangers.
This was originally reported in IP's 1990 50.59 annual report. This latest safety evaluation was performed to satisfy the prerequisites and limitations of this procedure which require an analysis of decay heat levels, the expected beat.up rate of the spent fuel i
ponis, and the time required to restore the systems to their normal configuration prior to each use of the procedure.
This evaluation showed that there was adequate heat removal _ capability using one FC heat exchanger during the third refueling outage (RF-3).
Conservative 11tnits are provided to assure sale operation.
PACE 6 DISABLE REACTOR WATER CLEANUP SYSTEM PROTECTIVE CONTROL FEATURES DURING A DIVISION I BUS OUTACE Docunent Evaluated: CPS Procedure 3509.01C001 R0 Log Ntaber: 92 0039 Two protective features of the reactor water cicanup system (RT) will be bypassed during a Division I bus outage. These protective features are designed to trip the pumps on closure of the outboard containment isolation valve or on low flow indication.
The purpose of this change is to allow continuous running of the RT pumps.
These features may be bypassed in plant operational modes 4 or 5 with no handling of irradiated fuel in secondary containment, no core alterations, and no operations with a potential for draining the reactor vessel.
Division 11 isolations and trips will continue to function during this outage; however, these isolations are not required to be operable during the operational modes in which this modification is implemented.
DISABLE AUTOMATIC TRIP OF REACTOR VATER CLEANUP PUMPS DURING A NUCLEAR SYSTEM PROTECTION SYSTEM DIVISION II BUS OUTACE Document Evaluated: CP5 Procedure 3509.01C002 R0 Log Nteber: 92 0038 The control system for reactor water cicanup system puinps is designed to trip the pumps if the inboard containment isolation valve closes.
This modification allows installation of a jumper to prevent the pump t rip when the control power is not available during a Division II bus outage.
The Division I interlock will still be operable. This modification is only allowed in operational modos 4 or 5 with no handling of irradiated fuel in secondary containment, no core alterations, and no operations with a potential for draining the reactor vessel.
TEMPORARY ILECTRICAL POWER TO FIRE PROTECTION PANELS DURINC A 4160 VOLT BUS OUTACE Docts..ent Evaluated: CPS Procedure 3514.010005 R0 log Nunber: 02 0036 This procedure allows the installation of a temporary power supply to fire protection panels while the 4160 volt bus 1A1 la taken out of service for muntenance.
If the temporary power source were to fail, a loss-of AC-power alarm will sound in the main control room and the power source will automatically transfer to a 24-hour battery backup.
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PACE 7 PROVIDE TEMPORARY POUER TO FIRE PROTECTION PANELS DURING A 480 VOLT SUBSTATION OUTAGE Document Evaluated: CPS Procekre 3514.010010 no tog Wuiber: 92 0037 This procedure allows the installation of a temporary power supply to fire protection panels during an outage of the 480 volt suustation 1A.
If the temporary power source were to fail, a loss of AC power alarm will sound in the rnain control room and the power source will automatically transfer to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery backup.
DIESEL GENERATOR OVERSPEED TRIP TEST t
Docunent tvaluated: CPS Procedure 3882.01 R1 Log Wuder: 92 0033 The Division I and 11 emergency diesel generator overspeed trip test proceduro-was revised to allcw tripping one of the engines while the other engine continues to operate.
This change allows testing the overapcod trip setpoint for each engine to ensuro both engines are in the required overspeed trip range.
Each engine is equipped with an overspeed trip. When the trip is l
actuated, crosatie logic also trips the other engine.
This revision provides for disabling the crosstic logic sc that the overspeed trip on each engine can be tested.
This test method was used at startup and was recommended by the vendor. The diesel generator will not be operable, per Technical Specifications, during this configuration.
CllANGE TO STANDBY LIQUID CONTROL' SYSTEM DRAIN PATil Dmunent Evaluated: CPS Procedure 9015.01 R31, POR 91 0411 Log Wunber 91 0108 The surveillance procedure for ensuring the operability of the standby liquid control system was revised to allow waste water to be drained to the containment building drains and treated in the liquid radwaste treatment system.
Tests have shown that boron levels following 11guld radwaste treatment system processing are below levels that would be of concern in the reactor water.
This change was evaluated against an event at another plant where boron was not removed by domineralizers prior to reaching the reactor ve s s,el. The liquid radwaste treatment consists of a two-step process of evaporation followed by domineralization.
In this process, it is highly unlikely that signiN cant amounts of baron will contaninate the reactor coolant.
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PACE 8 REACTOR CORE ISOLATION C00LINC (RCIC) OPERABILITY CHECKLIST PROCEDURE Docuient Evaluatedt tPs Procedsre 9054.01 227, 9054.010001 229 Log Eseer: 91 0097 The reactor core isolation cooling (RCIC) system surveillance procedure was revised to incorporate changes made in the inservice inspection prograin and the quick start section for RCIC turbine startup.
The changes made to the procedure allow the cycling of the RCIC pump minimum flow valve, IE51 F019, at 1000 psig discharge pressure rather than 300 psig.
This valve is designed to open at a differential pressure of 1400 psi. The turbine governor system is capable of controlling turbine speed and systern flow within limits with the 1E51-F019 valve open or closed.
BIOCIDE TREATMENT OF WASTE SLUDCE LINERS i
Docuvent tweluated, tPS Procedure tPS P 03 001 Log Ndier: 92 0025 This procedure provides a method to disinfect a waste sludge liner with a 0.5 percent concentration of glutaraldehyde prior to commencing the routine dowatering.
Sample analysis of liners has indicated the presence of biologically produced methane and carbon dioxide gas.
Following treatment, the biocide is removed as the liner is dewatered.
The removed liquid is distilled, filtered, and demineralized.
Then it is sampled to determine suitability for return to cycled condensate. The chemistry group has total organic carbon detection capability.
The glutaraldehyde content in the processed water is determined prior to return to the cycled condensate system.
The equipment necessary for biocide addition is compatible with the materials of the equipment in which it is used.
Continuous gas monitoring will be employed during biocide treatment as a precautionary measure.
SUPPLEMENTAL EVALUATION OF RADWASTE SOLIDIFICATION SYSTEM Docunent Evaluated: tP$ Procedure PCP 03 003 Log N daer: 91 0132 The CPS radwaste disposal process uses a vendor-supplied mobile dewatering and solidification system.
A safety evaluation,_ prepared in 1989, covered implementation of this process.
The safety evaluation was supplemented in 1991 to address the waste delivery system, the radwaste overhead bridge crane, the turbine building crane, the electrical supply, and tho' air supply.
This safety evaluation also included a realistic analysis of a liner drop event.
The analysis demonstrated that the limits of 10CFR20 would not be exceeded by this event.
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PAGE 9 REACTOR PRESSURE VESSEL (RTV) STEAM DRYER CRACK Docuwnt ivaluated: tt 1 89 01 162 R2 tog Wueer: 92 0066 During the first refueling outage (RF 1), a crack in a reactor vessel stvam dryer was discovered. At that time, the crack extended approximately 6 7/8 inch from the bottom end of a channel weld.
Inspections performed d; ring the second refueling outage (RF-2) showed that a 1/4-inch ligament at the bottom of the weld was cracked.
In addition, it was determined that the upper end of the crack had grown approximately 3/8 inch.
It was also judged that the separation between the channel plate and skirt had increased from that measured during RF-1.
Comparison between the RF 2 and RF 3 inspection results has shown that the crack growth rate has slowed.
The growth during cycle 3 was found to be only approximately 1/8 inch.
Based on a comparison of the RF-
- 1. RF-2 and RF 3 inspection results, it is not anticipated that growth beyond acceptable limits will occur during the fourth operating cycle.
REVISE USAR TO REFLECT ADMINISTRATIVE CHANCES, USE OF SINCLE-USE RESINS AND VX SYSTEM OPERATIONAL CHANGES.
Docunent Evolu ted: tt 1-90 03 087 RO tog Nurber 92 0088 The change reviscN the USAR to reflect current CPS practices for the use of single use resins.
The change also adds source term information to reflect the use of toss away resins and the analysis comparison to 10CFR20 limits.
+r regeneration and non regeneration modes of operation, the radiological consequences remain approxitnately equal to or less than the phase separator or the concentrate waste tanks.
CLOSE DIVISION II SilUTDOVN SERVICE VATER PUMP MINIMUM FLOV LINE CONTROL VALVE Docunent Evaluated: CO * ' i Di 020 R1 tog We ber: 92 0055 Valve ISX173B provides a minimum flow bypass of the division 2 residual heat removal (RilR) system heat exchanger for the shutdown service water (SX) pump.
Post maintenance testing during the second refueling outage indicated leakage past the valve seat. The valve could not be replaced during the RF-3 outage due to scheduling constraints.
The valve will remain closed until it can be replaced during RF 4.
The continuous flow through the R11R heat exchanger will prevent microbiological 1y induced corrosion (MIC).
The flow rate is below the velocity at which the heat exchanger tubes would be damaged. There will be no significant impact on service life since there is no heat load on the shell side of the heat exchangers during normal plant operation.
The pump operation
'was evaluated, and although the pump motor would be subjected to a higher starting current for a longer time, there is still adequate margin between the motor starting current and the over current trip relay setting. Also, the estimated amount of leakage through the valve (<106 'PM) will not impact the effectiveness of heat exchangers supplied with cooling water by SX.
PACE 10 SilIELDS SOIL SERVICE FACILITY EVALUATION Docunent Evaluated: CR 1 91 04412 R0 Log Wsler: 92 0087 USAR Sections 2.1, 2.2 r.nd 6.4 were revised to doctutient the analyals which evaluated the possibility of exposure of ruain cont.rol roorn personnel to hazardous chemicals in the event of a release of anhyorous ammonia f rom the Shields Soil Service Facilicy located in the vicinity of Clinton power Station.
No equipment, systems or parameters are affected by this charge.
RADVASTE EVAPORATOR BODY FOAM PROBE REMOVAL Doewent Evaluated: CR 1 91 07 016 R0 Log Nunter 91 0^*2 This change removed three evaporator vapor body foam probes and replaced them with blank ilanges. These probes and the associated electronics had previously been retired in plsce.
An cvaluation determined that the probes were deteriorating and damago could occur to system components if they were not reinoved.
The probes did not perform any control function.
The replacement bisnk flanges are equivalent to the flanges installed with the probes and will inaint ain the in egrity of the tanks.
TORQUE SWITCit SETPOINTS OF MOTOR OPERATED VALVES 1FF051, 1FP053 AND 1FP054 Docment Evaluated: tk 1 92 03 J36 R0 tog Wmtier: 92 0072 b
The purpose of this evaluation is to document the acceptance of the existing motor operators and torque switch setpoints for fire protection system containment isolation valves, 1FP051, IFP053, and IFP054 unt!1 completion of J
the NRC Ceneric Letter 89-19 evaluat. ion.
Design values for stroke times and flow rates will not be altered as a result of this condition.
Ealsting technical spncification requirements are still being rut and the valver wili function as intended to ineet the licensitig basis.
LCST PARTS FROM FUEL llANDLING GRAPPLE INTO Tile REACTOR VEs3EL Document Evaluated: CR 1 92 03 064 R0 tog morber: 92 0065 During the third refuelir.g outage, stainless steel fragments of an electrical connector were accidently drup wd into the reactor vessel during refueling activities.
The fragments were not recovered and are acsured lost in the reactor vessel.
The analysis report shows no potential interference with control rod operation >r with the core '.sw systea.
There is a very remote chance that the fragment will cause fuel c1cd fretting failure cfter prolonged operation.
This remote chance is from the fingments enrcring the bundle from either the upper or the lover tio plates.
The presence of the stainless steel fragmente in the core is not a concern since stainless steel is approved for use in the reactor, and therefore, is not a concern for potential adverso chemical reactions.
pAGE 11 DISCREPANCY BETWEEN Tile USAR AND Tile DESIGN CRITERIA POR THE FUEL POOL COOLINC (FC) IIEAT EXC!iANGERS Docuent tvaluated: te 1-92 03-07f. R0 tog Wueer: 92 0076 The plant configuration meets the design criteria requirements of having component coollrg water system (CC) flow on the shell side and fuel pt,o1 cooling and cicanup sy stem (FC) flow on the tube sido of the heat exchanger.
USAR Sections 9.1.3.2 and 9.1.3.1.2 were changed to reficct the correct configuration of the FC system interface with the CC system. The USAR was corrected to show that component cooling water flows on the shell side of the FC heat exchangers, and FC water flows through the tube side.
The plant configuration meets all the design criteria requirements.
LOST PAR'*S FROM QUARTZ LAMP LENS IN Tile REACTOR PRESSURE VESSEL (RPV)
Docunent [vetuatect tt 1-92+05 007 90 Log Nader: 92-0075 During the roasserbly of t.be RpV, glass particles f rom a quartz light lens were found in the reactor cavity grating area and on the RpV flange.
Several pieces, totaling approximately five square inches of the 1cus, were not reenvered and are assumed lost in the RpV.
The grating area around the RpV flange was flooded at the tirne of tho break.
The steam separators and the steam dryer were installed at the time of discovery which prevented the pieces Irom falling directly into the core region. This evaluation thornfore assumes that all of the missing fragments fell into the annular downcomer region of the reactor.
If the glass pieces had not fallen into the reactor, they would have entered the fuel pool cooling and cleanup (FC) system. The FC system is designed to remove suspended particles and impurities.
Therefore, it is expected any glass fragments entering that system would be removed.
The conclusion is that reactor operation is safe with the glass fragments remaining in the vessel.
The analysis performed shows that there la no concern for fuel bundle flow blockage since any glass fragments will be quickly reduced to small, sand-like grains by coolant flow. Theso grains will not inhibit flow through bundles, but will be dispersed in the reactor coolant ani eventually collected by the reactor water cleanup system.
Should any of the glass pieces migrate to the lower plenum and move to the core region, their area would be too small to cause fuel bu idle flow blockage.
Due to the small size and brittleness of the glass fragments, it is judged that they are incapable of causing laterference with control rod operatior. The prer.ence et theso glass fragments in the reactor is not a chemical h uard to the fuel or plant opetation because the glass will break un into small grains, and the oxide composition of these grains will remain chemically inert St reactor coolant temocrature.
Therefore, thera is..o potent tal for adverso chemical reaction.
l PAGE 12 INSTALL BARRIER f0 PROTICT DRYWELL PERSONNEL FROM RADIATION DUE TO DROPPED FUEL BUNDLE Docanent Evetuated: ICN 9730 Log kwtier: 92-0026 This change provides a temporary barrier to be installed in the reactor cavity during refueling outages.
The barrier concists of several platforms of metal r
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g ating which are mounted to the reactor head insulation bolts.
The barrier s y s t eru l-named m ovable guaranteed gecidental Icfueling drop phielding or
('.llMCARDS USAR Section 9.2 is being revised to reficct the REMCARDS installation. The design and installation of the RDiGARDS roodules have been analyzeJ f3r a seismic event and for the dynamic irrpart of a dropped fuel t.uitdl e During normal operations, the REMCARDS modules will be stored in the steam acparetor pool.
Storage of the modules in the pool was also analyzed fo; a wef mic event.
Storage of the modules in the separator pool will not decrease the amount of water available for an upper pool dumn since the modules will be stored belot the levW of the upper pool dump lines.
EMERGENCY CLASSIFICATION PROCEDURE Docusnt Evaluated: EPir EC 02, R3, ACN 4/3 t og ka4,e r : 92 0070 y
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This procedure (EPIP EC-02) was revised t o enhance ths recognitica and clas-i f ication of eine rgency condit ions. and also, to make emergency action levels consistent wi th other approved CPS doctunent s.
This change does not involve the design or operation of plant equipment or the operations training of personnel required to operate planc equipment.
This procedure only provides guidelines for classifying an accident.
The revision to this procedure meets the requirements of. 10CFRSO.$4q and 10CFRSO Appendix E EMEPGENCY OPERATING PROCEDURE REVISIONS Docuwnt Evaluated: E mergency Operating Pr ocedures ([0P4)
Log kmtwr: 92 0059 In response to an NRC Region III Notice of Violation and open items in NRC Inspection Report 50-461/91006, several changes were incorporated in the Clinton 7.0Ps and support procedures.
These changes corrected deficiencies identified in the inspection rep,rt, and were consistent with Revision 4 to the Emergency Procedure Guidelines RELOCATE EMERGENCY RESPONSE IIEADQUARTERS SUPPORT CENTER Docunent Evaluated: [ PIP HQ 01, R3, ACW 4/1 L og Nwter: 91-0091 IP han re'.ocated the emergency response lleadquarters Support Center to a new location at the IP Plaza in downtown Decatur.
This is a change to the CPS F nergency Plan which in section 13.3 of the USAR.
This change has no ittpact on plant operations and continues to meet the requirements of 10CFR50.S4(q) criteria.
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PAGE 13 CitANCE AUTHORIZED REPRESENTATIVE FOR CERTIFYING Tile ACTIVATION OF A LICENSED OPERATOR FROM VICE PRESIDEN7 TO DIRECTOR - PLANT OPERATIONS Doewent tveluated: NTD 2.13, R3, ACW 4/2 tog heter: 91 0079 10CTR55 53(f) requires that an authorized representative certify that the qualifications and status of a licensed individual, who has not maintained an active status, are current and valid prior to resuming license functions.
Previously, USAk Section 13.2.2.1.3.A required the Vice President to perform this function. This change requires the Director Plant Operations to approve the certification.
This change does not impact 10CFR$5.59 requalification.
SUPPLEMENTAL RELOAD LICENSING REPORT FOR RELOAD 3, CYCLE 4 toe w M tweluated: Report 23A?t44 Log hmber: 92 0064 Ceneral dicct ic report 23A7144 documents the results of reload licensing analyses for the next (fourth) operating cycle.
Included in this document are the plant conditions assumed in the analyses; fuel bundle types in the core and their proposed locations; applicabic margin improvement and operating flexibility options; and core and transient analysis results.
In a proprietary supplement, are maximum, averaSe planar, and linear heat generation rate (MAPLilGR) values as a function of average planar exposure for each unique axial region (lattice) of the new fuel bundles to be irradiated for the first time in the fourth operating cycle.
184 irradiated fuel bundles were replaced by bundles of a new type, CE8B.
Field alteration NBF010 documents the design changes and associated engineering reciews for using the CE8B fuel at CPS.
The new bt.ndles are standard, pre pressurized, barrict-clad, S lattice fuel bur.dles which will be described in a future update to G2 document NEDE-31152P, the fuel assembly description document that is referenced by the CE Standard Application for Reactor Fuel (CESTAR).
The CE8B fuel bundle design retains many of the features of the previously-loaded GE7B fu '.
Several changes have been made to accommodate higher fuel burnup, including:
increased helium pre pressurization; reduced clad to-pellet diametral gap with an increased pellet diameter; increased fuel density; cnd increased fission gas plenum volume above shortened active fuel columns in gadolinia bearing rods.
The first three changes will improve thermal conductivity in order to reduce fuel temperature and thereby reduce the rate of fission gas release from the fuel.
The increased fission gas plenum volume will accommodate the lower thermal conductivity of gadolinia-bearing fuel pellets.
Because of the potential for higher bundle-sverage enrichments and an associeted mote negative void coefficient of teactivity, CE8B incorporates'a high-flow upper tie-plate to reduce pressure drop in the two phase flow region of the bundle, thereby maintaining an adequate thermal-hydraulic stability margin.
In-order to improve the axial power distribution and cold shutdown reactivity margin, gadolinia is axially varied in-the bundle. Axially varying gadolinia was previously used_to improve the power distribution and reduce the inventory of shallow control blades.
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PACE 14
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The technical information-in report 23A7144 has been incorporated into the Clinton USAR by reference.
In addition, the cps Emergency Plan, which is referenced in Appendix D and section 13.3 of the USAR, is affected because the Emergency Classification Guide (Table 4 4a) is isopacted by a change in maximmn suberitical banked withdrawal position due to the loading of CE8B fuel.
For the sarne reason, the Emergency Operating Procedures, which are referenced in section 13.5.2.1.3 of the USAR, are also inspected.
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These changes are being inade to ineet cyc1r energy requirements at incree discharge burnups while operating in accordance with plant technical specifications.
Technical specification changes have already been made which established the Core Operating Litnits Report (COLR) and removed cycle-specific parameter limits from the technical specifications.
In addition, a request to change the technical specification bases to reference the COLR for the MAPIJlGR multiplier for single recirculation loop operation has been submitted to the NRC.
The core operating liraits established for the fourth operating cycle incet the requirements of the technical specifications since the CE8B fuel design has been incorporated into GESTAR.
ruel storage requirements in Technical Specification 5.6 are not impacted for CE8B fuel since the basis of this section is reactivity rather than enrichtaent. Technical Specification 5.3 states that "cach fuel rod shall have a nominal active fuel length of 150 inchen."
In CE8B fuel bundles, each fuel rod containing gadolinia has a reduced active fuel length in order to extend the length of the fission gas plenurn. This generic design change has been reviewed and approved by the NRC.
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PACE 15 ADLENDUM TO TiiE SUPPLEMENTAL RELOAD LICENSINC REPORT FOR RELOAD 3, CYCLE 4 Docment tvaluated: Report 23AT144 Log Wmters 92 0081 Safety evaluation 92 0064 documents the evaluation of reload licensing analyses for reload 3, cycle 4.
Included in this evaluation are the plant conditions assumed in the analyses; identified fuel bundle types in the core and their proposed locations; applicable margin improvement and operating flexibility optfor.s; and core and transient analysis results.
The Core Operating Limits Report is the cycle specific document which provides the operating limits for the following parameters:
the average planar linear heat generation rate; the core flow and power dependent minimum critical power ratio (MCPR); and the linear heat generation rate.
In accordance with Technical Specification 6,9.1.9, these limits are determined using previously reviewed and approved analytical methods, and are established so that all limits of the safety analyses are met.
IP was informed that the non applicability of the misoriented bundle accident to CPS may no longer be valid. This is a generic problem for all S lattice plants.
Calculations were performed for the change in critical power ratio (delta-CPR) for the misoriented bundle accident for CPS. The resulting deica-CPR of 0.13 for GE8B fuel exceeds that for both the 100*F loss of feedwater heating event and the rod withdrawal error.
Since these are the limiting events for cycle 4 operation, the MCPR operating limit may be impacted.
Proper orientation of fuel assemblies in the reactor core was verified during RF-3.
Several different visual indications of correct orientation ensure that any misoriented assembly would be readily identifiabic during core loading verification.
Vhile the generic issue is being discussed with the NRC, the Core Operating Limits Report was revised with a modified, flow dependent MCFR limit curve for the CE8B fuel to accommodate this increased delta CPR.
Specifically, the portion of the curve corresponding to rated and near rated core flow is being moved up to 1.20 from 1.18.
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PACE 16 0FFSITE DOSE CALCULATION MANUAL, REVISION 8 Document tvetuated: coca, sey. 6 tog wwters 91 0076 The Offsite Dose Calculation Manual (ODCM) provides the methodology to assure corepliance with the radioactive effluent done litnitations stated in 10CFR20, 10CF*50 Appendix A, 10CFR50 Appandix 1, and 40CFR190. Tl 3 'rirnary A
requirements for the CPS Radiological Environmental Monitos ng Progre (REMP) are also set forth in the ODCM.
The REMP conforms to the guidance of 10CFR$0 Appendix 1.1.
The ODCM was revised to delete the general surveillance requirement in Section 1.3.1 which stated that the combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
The revised surveillance interval will allow an extension not to exceed 25 percent of the specified surveillance interval.
t This change is consistent. with the Technical Specification surveillance requirement perforrnance requirement as amended in CPS Technical Specification Amendment No. 53, approved November 1, 1990.
The following changes were also made in Revision 8 to the ODCM:
Paragraph 2.2.1 was deleted and the criteria describing representative sampling for batch releases of radioactively contaminated water from CPS were added to Tabic 2.3-1 notations.
Figures 2.1-2. 2.5 1, 2.5-2, 2.5-3, and 3.3 1 were updated with editorial changes for case of reading or more accurato descriptions.
Section 2.5.2 was revised to provide clearer guidelines on the amount of radioactivity in liquid radwaste that may he stored in temporary liquid radwaste hold-up tanks. Tabic notation "g" for Table 3.4 1 was revised to clarify sampling requirements of charcoal and particulate on the station heating, ventilation and air conditioning and standby gas treatment system stacks.
In Table 3.9.2-1, note "1" was added to c1carly identify what is required to consider a particulate or iodine sampler operabic.
Thermoluminescent dosimeter sampling locations CL-?S, CL 96 and CL 97 were added to Figure 5.0 4 and designated as control locations. Requirement 2,a of Section 7.2 was revised to require total cont:Ircr volume vice container volume be reported for each class of solid waste shipped offsite during the report period in the Semiannual Radioactive Effluent Release Report per Appendix B of Regulatory Guide 1.21, i
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t PRE 17 TS;ndLISil AN OUTAGE PERSONNEL STACING AREA IN THE RADVASTE BUILDING AT Tile 762 FOOT ELEVATION Occunent Evolunted: Staging Arte Log Nadser: 92 0020 The 762 foot elevation of the radwaste building was utilized as a personnel staging area during the third refueling outage (RF 3).
This evaluation allowed the increase in combustible fire loading. USAR Appendix E - Fire Protection Evaluation Report, Section 3.7.4.3, was affected due to the increase in the fire load from office equipment and other appurtenances necessary for personnel occt'pancy.
There are no safety rel.ited cable trays or equipment located in this fire zone. The additional fire hazard is acceptable with additional administrative controls established per CPS Procedure 1893.01,
" Fire Protection Impairment Reporting".
TELEPil0NE SYSTEM UPCRADE Doewent Evolueted: USAR 1.2.8.17, 9.5.2.2, and 19.4 32 Log Natser: 92 0103 The inain solid-atate switchboard for the plant telephone systen was moved from the Service Building to the Nuc1 car Support Building. The telephone systems are powered frota a 12 W loop outside the plant protected area instead of a non-1E motor control center (HCC). A 52 volt DC battery provides backup power instead of a Division 1 power supply.
The USAR sections 1.2.2.8.17, 9.$.2.2 and Table 9.4-32 were revised to reflect these changes and also include that IP owns and maintains the telephone systems instead of a local phone company.
PROCESS DIACRAMS AND OTHER FIGURES Docunent Evaluated: USAR 1.7.4 tog Nu'6er: 92 0068 The USAR was clarified to say that process diagrams and other figures shown in the USAR, while representative of typical plant design and operations, may not be current, Current design basis documents and drawing should 'oe referred to when reviewing and evaluating the design.
It further states that the annual updates to the diagrams and figures will be provided when permanent corfiguration changes are made. This change is an administrative change.
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s PAGE 18 DOCUMENTATION OF POST LDCA HYDROGEN GENERATION Doctanent Evaluated: USAR 1. A, Regulatory Guide 1.7 Log Neber: 92 0058 USAR Section 1.8 was clarified to identify tnat specific analytical data a
pertaining to post LOCA hydrogen generation in the containment and drywell are subject to change.
The postulated post-LDCA hydrogen generation concentrations during the operational life of the plant are to be documented in design baseline calculations rather than in the USAR.
The specific data points and values in the USAR which were utilized lu, or were a result of, the hydrogen generation analysis will not be removed from the USAR. However, this information in the USAR is only representative of analytical methodology and results.
CPS procedures and baseline documentation ensute the limitation of post LDCA hydrogen concentrations inside the containment and drywell in accordance with Regulatory Guide 1.7.
USAR SECTION 1.8, REGULATORY CUIDE 1.33, Revision 2 Docunent Evaluated: USAR 1.8, Regulett,ry Guide 1.33 Log Nunber: 92 0042 USAR Section 1.8 was revised to take exception to the ANSI N18.7 requirement of biennial procedure review and identify programmatic controls that are equivalent to or better than the biennial review process.
The change was made to reduce the procedure review workload when the only requirement for this review is the biennial requirement.
USAR SECTION ?.1.2.2.5 Docunent Evaluated: USAR 3.1.2.2.5 Log Nw ber: 92 0093 The USAR Section 3.1 was revised to reflect that residual heat removal check valves 1E12-F050A/B and lE12-F053A/B are not equipped with position indication in the main control reom.
These valves perform the function of a third isolation barrier and are therefore not required to have open or closed indication. This is a documentation change only. No physical changes are being made to the plant.
No systems, equipment or parameters will be affected by this change.
TYPE "B" LEAK TESTING FOR ICIC HEAD SPRAY LINE PIPING FLANGED CONNECTION Docurnent Evaluated: USAR 6.2.6.2 Log Ntaber: 91 0044 The reactor core isolation cooling (RCIC) system _ piping to the reactor head is provided with a flanged connection to facilitate reactor pressure vessel head removal.
This change adds this connection to the USAR listing of items for which type "B" primary leakage testing is specified per 10CFR50, Appendix J.
This is a corrective. action required by Licensee Event Report (LER) 90 18.
This testing is performed when RCIC is not required to be operable.
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PACE 19 RESTRICT TESTING OF UNTESTED ISLANDS IN NUCLEAR SYSTEM PROTECTION SYSTEM TO THOSE AFFECTING FUNCTIONAL LOGIC s
Document Evaluated: U$AR 7.2.1.1.4.8 Log Nunters 92-004$
The nuclear system protection system (NSPS) is provided with a self test system (STS), which automati ally tests the NSPS logic continuously; however, there are some components of the NSPS logic circuits which are not tested by the STS.
The untested portions are referred to as untested islands (UTIS).
CPS committed to the NRC to perform mcnual testing of the UTIS.
Manual testing _ involves removing the circuit cards from the NSPS panels.
Removing the circuit cards has resulted in greater impact on equipment and c;erations than those failures that the tests are designed to avert. There have been three Licensee Event Reports (LERs88-009, 89-012, and 90 015) caused by removing or inserting circuit cards in the NSPS panels since testing began in 1987. A failure modes and effects analysis (FMEA) was performed which indicated that continued testing of several UTI types is not necessary. For these UTls, it has been determined that those failure modes which are not detectable by STS either do not inhibit the safety function of the NSPS logic or are detectable during the performance of system or instrumentation surveillance tests currently required by the CPS Technical Specifications.
The types of UTIS no longer subject to manual testing are:
self test coupling capacitors, logic seal-in or latching circuits (20 millisecond delay and 250 millisecond delay), and power-on initialization circuits.
'ecause the above UTIS are no longer being manually tested, USAR Section 7.2.1.1.4.8 is being revised to state that special manual tests are only required to detect those failures which could prevent the NSPS from performing its intended safety functions.
Since UTI' testing is not addressed by the technical specifications, this change does not alter any technical specification.
Testing needed to meet the requirement = of logic system functional tests will continue to be performed as required.
Some of the failure modes of these UTIS could result in' spurious operation of safety equipment.
However, elimination of the requirement to periodically remove the associated circuit cards has the potential to improve the performance of the associated equipment by eliminating the potential for spurious trips while removing or inserting the cards in the NSPS panel.
CLARIFY THE RATED VOLTAGE FOR MOTORS OF MOVS Dxument Evaluated: USAR 8.3 10 Log Nurser: 92 0061 In 1991, an NRC inspection evaluated the methodology at CPS for evaluating the l
capability of motor operated valves (MOVs) to achieve the required thrust.
The results of this inspection are dor.rmented in NRC Report 50 461/91019.
The inspection evaluated the actions being taken in response to Generic Letter 89-10 (Safety Related Motor Operated Valve Testing and Surveillance).
The USAR was clarified to state that continuous duty motors are capable of producing full load torque at 75% of rated.voltege, and Class 1E motors for motor--
operated valves are designed to open or close the valve against specified differential pressure, without damage, at 90% of rated voltage.
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PAGE 20 REVISE USAR TO RECOGNIZE THE POTENTIAL FOR NEUTRON ACTIVATION OF CORROSION INilIBITOR CHEMICALS IN THE PLANT CilILLED VATER AND DRYVELL CilILLED WATER SYSTEMS Docunent Evaluated: U$AR 9.2.8 Log su6er: 91 0105 The USAR was revised to acknowledge the presence of low radioactivi,y levels (10E 7 to 10E 6 microCi/ml.) within the drywell v-ntilation (VP) and plant chilled water (WO) systems and to institute requirenients for the review of any changes to the amount or type of corrosion inhibitor added to these systems.
The radioactivity levels resulted frem neutron activation of the corrosion inhibitor which was adde /. to the WO and VP systems.
The USAR is being changed to acknowledge that the corrosion inhibitor in the WO and VP has been neutron activated inside the drywell.
The radiation levels resulting fr9m the activation do not pose a concern to individuals workin6 on these systems or a contamination problem in the case of spills or leaks.
The proposed change to the USAR does not introduce an em eviewed saf ety question, increase the probability or consequences of an accident or malfunction, or introduce a new type of accident or malfunction.
DELETE OPERATIONAL CRITERIA FOR Tile SERVICE BUILDING VENTILATION SYSTEM FROM Tile USAR Docunent fvaluated: USAR 9.4.12.1.2.6 tog kueer: 91-0098 The service buildin6 ventilation systern descrition in USAR Section 9.4.12 was revised to remove the statement "The system maintains the offices at 75'F 12*r and 45 15% relative humidity (Rll) in the stnamei and 35 15% Ril in winter".
This change allows the personnel working in t.he office ::reas of the service building to adjust the room temperatures for their own personal comfort.
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i PACE 21 DIVISION III DIESEL CENERATOR AIR COMPRESSOR TACOUT Docunent Evaluated: USAR 9.5.6.2 Log Ww ber: 92 0091 The diesel-driven air compressor (IDC03CA) has been removed from service for extended periods of time because of check valvo problems.
At tho time of the safety evaluation, the air receiver inlet check valve (ICC172) to the air receiver associated with the diesel would not open.
Normally, when either the motor driven air compressor (1DC03CB) or the diesel-driven compressor starts, both air receivers (IDG06TA,B) arc pressurized equally.
In the present configuration, the ai receivers require manual equalization to maintain the pressure in both receivers. This is done by momentarily opening (for approximately 30 seconds' the air receiver cross connect valvo (1DC631) per CPS Procedure 3506,01.
The equalization process ensures that both air receivers maintain their safety design basis capability of five successive starts without recharging the air receivers.
The USAR states that the Division 111 dieset generator has two air supply subsystems.
Each subsystem consists of one air receiver connected to one starting motor train, and each wir receiver is charged by an individual air compressor.
USAR Section 9.5.6.2 will be changed to clarify that the compressors are connected to a common air dryer and the air *cceivera can be charged by either or both air conpressors. Also, the equaliz..lon valve may be used to manually equalize the pressure between the air ceceivers.
INCREASE VOLUME OF RESIN AND ALLOW RATIOS OrilER TilAN 1:1 FOR CATION / ANION USED IN LIQUID RADVASTE DEMINERALIZERS Docuwnt Evaluated: USAR 11.2.1 2 Log Nm ber: 92-0001 USAR Section 11.2.1.2 was revised to state that a total of 90 cubic fact of anion and cation resin will be used in each liquid radwaste treatment system demineralizer rather than stating a specific ratio.
CPS does not regenerato resin, therefore, tiie change will increase the ratio of anion to cation resin resulting in a decrease in the ionic concentration in the waste stream.
This change will not increase the concentration of radioactive source terms-above that used in the analysis of the ruptured concentrated waste tank because the source terms for each radwaste waste stream are traced to ensure they remain below those assumed in the accident analysis.
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REVISION TO RADVASTE CONTAINER DROP ACCIDENT ANALYSIS I
Doctsnent Evstuated: USAR 11.4.1.6 Log Neber: 92 0056 The CPS mobile radwaste solidification system was revised to allow waste to be dewatered or solidified in.large volume waste containers. Operationally, this results in fewer lifts made with the radwaste crane,.but the loads are heavier.
Some loads must be handled with the turbine building crane instead of the radwaste crane.
Section 11.4 of the USAR was revised to reflect this change. An analysis of a ruptured liner was performed.
It demonstrated that the limits of 10CFR20 are not exceeded. Also, the radionuclide concentrations which may be anticipated to occur throughout plant operation including fuel failure were used-in the calculation in order to determine the maximum concentrations which approach the 10CFR20 limits.
These concentrations are periodica 1y monitored by CPS 10CFR61 compliance program.
These radiot.uclide concentrations are also being incorporated for reference into USAR Table 12.2-12.
LIQUID RADVASTE TANK FAILURE ANALYSIS Doc snent Evaluated: USAR 15.7.3 Log Ntsnber: 92 0035 USAR Section 15.7.3 provides an analysis of a postulated radioactive release due to a liquid radwaste tank failure.
The original calculation was revised to base the cnalysis on the rupture of the concentrated waste tank instead of the phase separator tank.
The concentrat-d was_2 tank was chosen for analysis because it has the highest average radioactivity that could become airborne.
Although the radionuclide inventory for the phase separator tank is greater than that for the concentrated waste tank, the radio;'uclides in the p! ase l
separator are f.onically bound to the phase separator resins, and therefore, less likely to go airborne than those in the concentrated waste tank.
The methodology used to perform the calculation for whole body and inhalation doses at the exclusion area boundary and in low population zones reflects a
- n. ore realistic, but still conservative calculation, The methodology used in these calculations is acceptable for dose calculations.
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- REVISED CRITERIA FOR FIRE-k.'.TED SEALS IN ELECTRICAL CONDUIT Doctamt Evsluateds' USAR APP E 3.1.2.2.10, 9.5.1.1.d Los utaber: 91 0101 This change revised CPS's position, regarding periodic inspection and installation of internal fire-rated conduit seals, which states that all penetrations through fire barriers such as mechanical piping, electrical conduit and cable trays will be sealed with fire-rated seals. USAR.Section 9.5.1.1-d was also revised to include reference to the " Conduit Fire Protection Research Program", Final Report, which provides the technical justification for revising internal conduit fire seal installation and inspection practices.
The report was accepted by the NRC in a letter, dated October. 23, 1989 titled,
" Review of Draft Safety Evaluation of Conduit Fire Seal Topical Report for Proprietary Contract," and was substantiated with a safety evaluation and technical evaluation report.
The NRC staff evaluated the research program in tP' safety evaluation and technical evaluation report and concluded there is justification for the revised conduit sealing criteria.
The report provides guid611nes regarding sealing of conduits penetrating fire barriers.
The guidelines are consistent with the proposed change.
Other justification for acceptability of the change includes:
past Quality Assurance (QA) and Quality Control (QC) inspections of internal conduit seal material installation during the construction-phase, and continued QA material installation inspections during the operational phase; maintenance work control process; engineering (configuration control) design change and associated design review process; limited accessibility and small cross-sectional area of material exposed; seal material qualification for the plant life; and limited possibility for propagation of fire from one fire area to another.
CORRECT DISCREPANCIES IN RELIEF VALVE DISCHARGE LOCATIONS Document Evaluated: USAd F3.6-1, F6,2-133, F9.1-4 Log Ntaber: 91-0071 USAR Figures 3.6-1 and 9.1-4 were revised to correct the location for the discharge of relief valves 1FC085A and-lFC08FB, which discharge to the
-suppression pool via the suppression pool makeup-(SM) system stean separator storage pool dump lines. Also,JUSAR Figures 3.6-1 and 6.2-133 were revised to correct the discrepancy for the location of relief valve 1SM003A. As the actual-relief valve discharge locations are adequate there is no increase in the probability or consequences of any malfunction or accident analyzed in the USAR.
e.. + Po PAGE 24 REACTOR VATER CLEANUP SYSTEM PROCESS DIACRAMS Docwent Evaluated: USAR F5.4-17, $1 & 2 Log Nu ter: 92-0067 USAR Figure 5.4-17 was revised to reflect the current design maximum flow rate for the reactor water cleanup system of 450,000 pounds per hour (907 GPM at 110*F shell inlet temperature) is acceptable during startup and shutdown conditions. The changes were the result of concerns that t.& operation of the reactor water cleanup system in the recirculation mode would provide higher reactor water temperature to the feedwater system piping, which has been analyzed for a lower operating temperature.
This recirculation mode of operation precludes thermal stratification in the bottom of the vessel when the reactor recirculation pumps are not operating.. Administrative controls were added to the operating procedure to ensure that the feedwater piping is not heated up above 435'F.
AS-BUILT BRF.ATHING AIR SYSTEM PIPING AND INSTRUMENTATION DIACRAM (P&ID)
Docment Evaluated: USAR F9.3 3 Log NJter: 90-0071, R1 This USAR change corrected Figure 9.3-3 to as-built conditions for the breathing air (RA) system.
The changes consisted of correcting a pipe diameter and a reducer location.
This change did not affect the system components and did r't increase the consequences or probability of occurrence of RA failures evaluated in the USAR, CORRECT SEISMIC CLASSIFICATION OF THE REACTOR VATER CLEANUP PUMPS Docunent Evaluated: USAR T3.2-1 Log Nuter: 91 0114 USAR Table 3.2-1 was revised to change the seismic category of the reactor water cleanup (RT) pumps from seismic category I to N/A.
Although the RT pumps are not seismic category I, the nuclear steam supply system design i
l criteria and ASME Section III require the consideration of seismic loading in the design of the pump.
BATTERY LOAD REQUIREMENT CALCULATIONS Document Evaluated: USAR T8.3-8, 9, 10, & 11 Log Nunber 92-0090 USAR Tables 8.3-8 8.3-9, 8.3-10 and 8.3-11 for Division 1, 2, 3, and 4 were changed ta reflect a change to the 125 volt battery load amperage requirements. These values were adjusted based on a design review of battery load calculations. The calculation reviews were made to provide greater detail and documentation support for the analyzed battery loads.
The revised loads remain within the design capability of the batteries.
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REVISE DIVISION III BATTERY LOADS Doctanent tystuated: USAR 18.3 10 tog wtsster: 91-00T3' The values in USAR Table 8.3-10, for the Divison III 12 volt battery load requirements, are being revised an a result of a general design review of the battery loads.
The major change was the incorporation of the starting current value for the Division III diesel-engine-powered air compressor starter motor. The reviued loads are within the design capacity of the battery. A change to Technical Specification 4.8.2.1.d.2.C is being processed but is not required prior to implementation of this USAR change since the changes are still in compliance with the Technical Specifications.
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