ML20069H970
| ML20069H970 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/08/1994 |
| From: | Scott A COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AMS-94-017, AMS-94-17, NUDOCS 9406140120 | |
| Download: ML20069H970 (190) | |
Text
Commonwecith Edison Quad Cities Nuclear Power Station 22710 206 Avenue North Cordova,lilinois 61242 Telephone 309/654-2241 AMS-94-017 June 8, 1994 U.
S.
Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555
SUBJECT:
Quad Cities Nuclear Station Units 1 and 2 Changes, Tests, and Experiments Completed NRC Docket Nos. 50-254 and 50-265 Enclosed please find a listing of those facility and procedure changes, tests, and experiments requiring safety evaluations completed during the month of May, 1994, for Quad-Cities Station Units 1 and 2, DPR-29 and DPR-30.
A summary of the safety evaluations are being reported in compliance with 10CFR50.59 and 10CFR50.71(e).
Respectfully, Comed Quad-Cities Nuclear Power Station d Y (f
i Anthony M. Scott System Engineering Supervisor AMS/dak Enclosure i
cc:
J. Martin, Regional Administrator C.
Miller, Senior Resident Inspector i
SAIET M RC.L R V '
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SE-94-37 Software Activity Request #991 DESCRIPTION:
The Rod Worth Minimizer (RWM) software was upgraded to include two new features.
First, a select block was added to the previous rod blocks. This causes a rod block upon selection of an out-of sequence control rod.
This select block function included an on/off ' toggle' on the RWM touch screen.
The other change is the provision of an insert block signal in Rod Exercise mode as soon as a control rod is moved in one notch.
In addition, a rod block was applied to all other rods at this time.
The blocks are removed when the current rod is withdrawn to its original position.
Previously the RWM applied a rod block only after a rod had traveled more than one notch past its target position.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assuned to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Rod Drop Accident UFSAR SFCTION 15.4.10 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a i
different type than any previously evaluated in the UFSAR is not created because the changes that will be made are to the RWM software only.
No modifications will be made to the RWM computers, nor will any interactions with other systems be changed.
The blocks that are being added will serve as an additional barrier to control rod mispositionings and all of the current rod blocks will be retained.
Because no other systems will be affected, there will be no adverse system interactions or accidents created in other systems.
TICIOP3\\SAILTY\\94MAY.RPT
i SE-94-37 CONTD J
In addition, because the RWM itself will not be altered, the failure modes will remain the same as before the software upgrade.
As a result, there will be no new type of RWM malfunction not evaluated in the UFSAR.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
TIrHOl%5AITTY\\94MAY.RFr
SE-94-38 QCOS 2300-1 Rev. 6 DESCRIPTION:
This procedure was revised to have the operator verify lube oil temperature indicating switch setpoints are set to their proper value as listed in tha procedure.
Also, the HPL7 pump is verified to be filled and vented locally prior rolling the HPCI turbine.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
1 The accidents which meet these criteria are listed below:
Small Break LOCA UFSAR SECTION 15.6.4, 15.6.5 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a i
different type than any previously evaluated in the UFSAR is not created because verifying lube oil temperature indicating switch setpoints prior to rolling the HPCI turbine doesn't adversely impact systems or functions so as to create an accident of a type different from those previously evaluated in the UFSAR.
This procedure change will only aid in preventing possible damage to the HPCI turbine and thus decreasing the probability of any accident.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
TiiCHOITSAMTYW4MAY.RIT
SE-94-039 One time extension to 67 day Safe Shut Down Administrative DESCRIPTION:
This change to the fire protection program has two parts:
allow the extension of the-67 day safe shutdown ATR for safe shutdown path B by 24 days.
establish additional compensatory measures.
BAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Appendix R Fire as described in the Fire Hazards Analysis For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this is an administrative change to the procedure and therefore dccs not effect equipment operation.
The type of accident that could occur (Appendix R Fire) has already been evaluated.
The stations approved program addresses the consequences of this fire.
No new accident types will be created by this change.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
i HCHOl'33An'.lY\\94MAY.RFT
SE-94-042 Temp Alt 94-2-32 Disabling Personnel Interlock on U2 DESCRIPTION:
This Safety Evaluation made changes to the Unit 2 Primary Containment Air Lock Doors and Air Lock Mechanism.
The following is a description of the changes:
The interlocks that prevent more than one door open at a time have been defeated and are being controlled as per Technical Specifications 3.7.A.7.c.
A valve that is connected to the interlock mechanism that equalizes pressure between the Drywell and volume between the interlock doors has been gagged in the CLOSED Position.
The "Stongbacks" have been left installed to ensure the Drywell side door is aligned and seated properly.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine cach accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Decrease in Reactor Coolant Inventory UFSAR SECTION 15.6 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
11CHOPMSAFLTYW4MAY.WT
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SE-94-042 CONTD l
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the following changes have been made to the Unit 2 Personnel Interlock Doors for Primary l
Containment.
j The interlocks that prevent more than one door open at a time have been defeated and are being controlled as i
per Technical Specifications 3.7.A.7.b and c.
One interlock door closed will ensure the integrity of Primary Containment.
Administrative controls as required by TS, are being implemented that will ensure only one door is open at a time.
i A valve that is connected to the interlock mechanism I
that equalizes pressure between the Drywell and volume-j between the air lock doors has been gagged in the 6
CLOSED Position.
The air lock doors are being leak rate tested to verify that there is no leakage or acceptable leakage out of the air locks.
The valve that communicates the Drywell volume and air lock volume has been gagged in the CLOSED position to ensure that during a seismic event and/or a decrease in reactor coolant that the valve will remain in the CLOSED (as tested) condition.
l The "Strongbacks" have been left installed.
-l The strongbacks have been installed and will be left on the Drywell air lock door.
The strong back is a series of l
structural steel that bolts onto the Drywell door and secures the door in the closed position for leak rata testing.
The strongbacks have been previously evaluated and found acceptable for operation.
Based on the above information and the fact that the Technical Specification LCO is being implemented the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR is not created.
3.
The margin of safety, as defined in the basis for any l
Technical Specification, is not reduced because the LCO t
requirements for an air lock door inoperable and the air lock interlock mechanism inoperable.
Therefore, the Technical Specifications will be meet.
This will ensure the margin to safety will be maintained.
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SE-94-043 SESR #4-2156 DESCRIPTION:
The installation of a larger U-bolt (5/8" versus 1/2") on pipe support M-987D-75 because of increased weight of a parts upgrade of the Standby Liquid Control (SBLC)
Accumulators for Unit 1.
The replacement accumulators evaluation is ME-93-0541-00, Revision 2.
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SAFETY EVALUATION
SUMMARY
i 1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implic 'tly assumed to function during or _
l after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Anticipated Transient Without SCRAM UFSAR SECTION 9.3.5.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or i
malfunction of equipment important to safety as previously evaluated in the UFSAR.
j 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the larger U-bolt does not create the possibility of an accident or malfunction of a type i
different from those evaluated in the UFSAR.
The larger U-bolt functions in the same manner as the smaller U-bolt it is replacing.
The higher weight of replacement SLC accumulators requires the larger U-bolt to maintain the i
system seismically.
Sizing of the U-bolt has been evaluated by seismic calculation (SESR 4-2156).
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because SESR 4-2156 evaluated the increase in weight of the replacement accumulator and the replacement of a larger U-bolt on pipe support M-987D-75.
The evaluation determined that these changes were within the design loadings of the SBLC system.
I 11rHOP3GAH,TYW4MAY.RIT l
M04-0-90-003 CRD Repair Room A/C Installation DESCRIPTION:
Provided cooling for the CRD Repair Room.
This modification installed an air cooled condenser located outside the room and an air handling unit located inside the room.
Electrical power is supplied from a GE MCC which replaced the existing Westinghouse MCC 42R-2-1.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or J
malfunction of equipment important to safety as previously j
evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this modification will install an air conditioning system for the CRD Repair Room.
The air handling unit, plenum, return grill and thermostat will be installed inside the CRD repair ante room.
The air cooled j
condenser will be installed outside the CRD Repair Room.
This equipment will be powered from a 480V GE MCC which will replace the existing Westinghouse MCC 42R-2-1.
Operation of the CRD Repair Room A/C System will offset the constant addition of heat incurred during maintenance on control rod drives.
The design includes locally mounted disconnect switches for periods when this equipment will not be required.
TECHOP33AfrIY\\94MAY.RPT
M04-0-90-003 CONTD Possible new failure modes or unacceptable conditions include:
1.
Electrical failures in the new equipment.
2.
Leaks in the new refrigerant Jines.
3.
Failures in the modified block walls.
4.
Spread of contamination.
Possible impact of the above failures during all operating modes are:
1.
The eleccrical requirements for this modification include the installation of properly sized breakers in non-safety related MCC 42R-2-1 to protect existing plant electrical equipment from any faults which may occur in the new HVAC equipment.
MCC 42R-2-1 receives electrical power from non-safety related transformer T42R-2.
The only loads on MCC 42R-2-1 will be the CRD Repair Room HVAC System.
Therefore, a fault in the new electrical equipment will result in the tripping of breakers in MCC 42R-2-1 which will have no impact on any other plant equipment.
2.
A leak in the refrigerant lines installed by this modification would result in the release of refrigerant-22 into the Unit 1 Reactor Building.
The Reactor Building Ventilation System, designed to produce a negative differential pressure, evacuates the Reactor Building at a rate of approximately 1 free volume / hour.
Therefore, leakage of refrigerant into the Reactor Building free volume would have no credible impact from a human safety standpoint and have no impact on equipment operation.
3.
The structural requirements for this modification include design changes to the west (blocking-in an existing louver opening) and north (installation of electrical supply and refrigerant supply and return lines) block walls.
As part of the designer's walkdown, it was identified that no safety related equipment was attached to these two block walls.
The actual design will require structural changes meet the seismic 2-over-1 criteria but, if a failure of the wall were to occur, no safety related equipment would be affected.
TIDIOP3\\SAFL'IY\\94MAY.RPT
l M04-0-90-003 CONTD 4.
Increased local air flow from the air handling unit could result in unacceptable spread of contamination.
The location of the air handling unit inside the ante room instead of the CRD Repair Room provides the highest air flow in the area of least contamination to prevent an unacceptable airborne contamination problem.
Blocking-in the louver opening seals the ante room to prevent the spread of contamination to an uncontrolled area.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
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DCR 4-93-205 1
DESCRIPTION:
j The torus level indicatir,n was found to be in error and the source was traced to the faulty Narrow Range Level Transmitter (LT-001-1602-9).
NWR Q08293 replaced the Rosemount Model 1151DP3B12 (obsolete designation) with a Rosemount Model 1151DP3G12M1B1.
This model includes the mounting bracket which was previously ordered separately and an optional integral meter.
l The Reactor Building Exhaust Fan 2C was auto-tripping and I
the source was traced to the setpoint of differential pressure switch (DPS-002-5741-261C) being at the low end of the switch's range.
NWR Q08212 replaced the Dwyer Model 1821-2 with Dwyer Model 1823-1.
This model has a range of 0.3" to 1.0" water column (WC) whereas the previous model's range was 0.5" to 2.0" WC.
The setpoint of 0.5" WC remains unchanged.
DCR 4-93-205 updated the appropriate data sheets to reflect these changes.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Internal Flood Measures UFSAR SECTION 3.4.1.2 LOCA UFSAR SECTION 7.5 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
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DCR 4-93-205 CONTD 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the instrument model changes will not affect the function or operation of the systems since the replacement instruments function the same as the-original instruments.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
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DCR 4-94-046 DESCRIPTION:
This DCR revised the Master Equipment List (MEL) and.
selected drawings to incorporate the results of Component Classification (CC) of the Standby Gas Treatment (SBGT) l System.
As part of this DCR, 1) no physical change was made to any plant structure, system equipment or component and 2) some components were upgraded from NSR to SR because they are required for the SBGT system to perform its SR function (Secondary Containment Radioactive Effluent Control).
Documentation specifically addressing these changes is included in Component Classification Binder #CC-QC009.
The CC program is an ongoing controlled program that is i
supervised by Station Engineering.
I SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
1 The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
t The accidents which meet these criteria are listed below:
Break in Reactor Coolant i
Pressure Boundary Instrument Line Outside Containment UFSAR SECTION 15.6.2 Loss of Coolant Accidents Resulting from Piping Breaks Inside Containment UFSAR SECTION 15.6.5 Design Basis Fuel Handling Accidents Inside Containment and Spent Fuel Storage Buildings UFSAR SECTION-15.7.2 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
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i DCR 4-94-046 CONTD 2.
The possibility for an accident or malfunction of a j
different type than any previously evaluated in the UFSAR is not created because this DCR does not involve any physical changes to plant systems, structures, equipment or components.
The Component Classification (CC) process for the SBGT system identified the operating mode for each component in the system and also identified that component's role in accomplishing the SBGT system safety function.
The CC process also considered all applicable accidents analyzed in the SAR and all potential equipment or component malfunctions.
The CC process provides assurance that the changes made by this DCR do not affect any existing accidents analyzed in the SAR and do not create any new accidents.
The SBGT system CC process is documented in the SBGT system CC binder.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
i TirHOP3dA11Tn94MAY.RPT
DCR 4-94-055 DESCRIPTION:
The implemented change incorporates the actual location of pressure test point connections for the condensate booster pump discharge piping on Unit 1 Piping and Instrumentation Diagram (P&ID); and incorporate the addition of pressure test point connection for the condensate booster pump discharge piping on Unit 2 P&ID.
These Unit 1 and Unit 2 P&ID as-built changes reflect the original designed and installed conditions.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Loss of Normal AC Power UFSAR SECTION 15.8.2 Loss of Normal Feedwater Flow UFSAR SECTION 15.8.3 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a j
different type than any previously evaluated in the UFSAR is i
not created because the change as described does not cause a functional change in the system or its interaction with other plant systems.
It does not alter any physical parameters or process variables of the plant.
Due to the nature of the change, there are no new inherent failure modes introduced to the system and the change does not add any new components or process routes.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefare, the safety margin is not reduced.
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DCR 4-94-064 DESCRIPTION:
Schematic Diagrams 4E-1351B, Sheet 2; 4E-2345, Sheet 1; 4E-2345, Sheet 2; 4E-2430, Sheet 2; and 4E-2430, Sheet 4: These drawings update cross references and descriptions on relays and control contacts to more accurately reflect the installed conditions.
Piping Diagram M-84, Sheet 1:
This drawing revised the Equipment Piece Number (EPN) for the Unit 2A Off-Gas Filter Outlet Valve from 2-5499-55 to 2-5499-51.
This change was made to match the configuration and numbering of the Unit i valve.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Loss of auxiliary power UFSAR SECTION 8.3.1 Power bus loss of voltage UFSAR SECTION 8.3.1 Failure of one diesel generator to start UFSAR SECTION 8.3.1.6.4 Load rejection without bypass UFSAR SECTION 15.2.2.1 Load rejection with bypass (Loss of electrical load)
UFSAR SECTION 15.2.2.2 Loss of Coolant UFSAR SECTION 15.6.2, 15.6.5 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in tne UFSAR.
TIrHOP3GAH3Y.94MAY.RIT
DCR 4-94-064 CONTD 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because no new accident scenarios are created by this DCR.
The function of the Core Spray, Diesel generator and Off-Gas Systems and their ability to operate are unchanged.
This DCR will not adversely impact systems or functions nor will the possibility of an accident malfunction be created that is different from those previously evaluated in the SAR.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
U! CHOP 3\\SAR'.IY\\94hMY.RIT
SE-93-109 M04-1(2)-89-115 Work Packages QO2824, QO2825 DESCRIPTION:
The work performed under this package calibrated two 0-100 psi pressure indicators (SI/208039) and a flow switch
]
(SI#699224) prior to installation of modification M04-1(2)-
89-115 (Modification of the service water radiation monitoring system sample delivery piping).
The pressure indicators (PIs) will be used to ensure the sample stream eductor is operating properly.
Under this package, the PIs will be used to gather system sample pressures while j
throttling the two glove valves on either side of the eductor.
Also, during this test, a flow indicator was installed to give flow indications.
This information was used to determine proper system operating pressures.
The flow indicator was then removed and the flow switch was installed.
The low flow setpoint was then verified.
If
'j erratic indication occured during performance of the i
traveler, individual instrument calibrations can be performed.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
]
The changed structure, system or component is explicitly or implicitly assumed to function during or I
after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
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SE-93-109 CONTD 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because installation and testing of this equipment cannot cause any plant accident or transient not described within the UFSAR.
The installation does not alter the interconnecting systems so as to create abnormal lineups or operating modes.
The installation will be passive with respect to the potential to initiate a different type of accident.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
k 11 CHOP 3\\SATTTYW4MAY.Rf7
SE-93-112 M04-1(2)-89-115 Service Water Radiation Monitor DESCRIPTION:
The work to be performed under these packages demolished the existing Service Water Radiation Monitor (SWRM) sample delivery system (receiver tank, pump, all associated piping and valves) and installed a new, eductor driven system powered by domestic water.
The sample system inlet isolation valve was replaced and the service water return header was open through a 1-1/2" pipe to the turbine i
building 595' level during the replacement.
This valve acts as the isolation point for further installation work.
Domestic water was isolated for installation of the back flow preventer.
All other items (skid, detector, eductor) were then installed.
A flow indicator was installed to facilitate Instrument Maintenance work and testing on the flow switch and pressure gauges.
The indicator was removed and replaced with the switch.
A leak test was then performed.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is I
explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
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SE-93-ll2 CONTD 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this work only interfaces with the domestic water system and the service water system.
Both interfaces are mechanical only.
No other SSC will be impacted by the scope of this work.
The worst case scenario would involve a failure of the installed isolation valve on the service water return header.
This would lead to leakage onto the turbine building first floor.
But, this leakage will not be of greater magnitude than the capability to remove water by the floor drain system.
Therefore, this event will not result in flooding.
No other SSC will be adversely impacted so as to create a new UFSR accident or transient.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
TIETIOl'3\\5AMTY\\94MAY.107
E04-1-93-094 Replace Turbine Rotor Unstacking Transformer with Dry Type DEMCRIPTION:
The subject exempt change replaced an oil-filled 1 MVA transformer with a dry-type 500 KVA transformer on elevation 639' of the Unit 1 Turbine Building.
The existing wet pipe system was demolished.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because T42R-5A receives power from the 13.8 KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8 kv system is not used for plant equipment.
Therefore, this transformer will not electrically affect operation of plant equipment.
Per the Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
The replacement of the oil-type transformer with a dry-type one results in no new accident type.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
TifHOPMSARTY\\94MAY.RPT
E04-1-93-245 i
Install Welding Receptacles on Turbine Shield Wall DESCRIPTION:
The subject exempt change replaced the existing panel with a 10 circuit distribution panel and installed six 60 amp welding receptacles powered from this new panel.
Five receptacles were mounted on the outside of the turbine shield wall west of the new circuit panel.
A sixth was installed on the inside of the turbine shield wall.
This new configuration provides a safer and more efficient means for providing power on the turbine deck.
i SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increcse the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
l 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the subject design change will not result in changed operation of the existing panels.
l Therefore, no new accident that has not been previously analyzed will be created.
i 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
UnionuMrnesbMY.MT I
r
E04-1-93-325 UAT Change-out Concrete Work DESCRIPTION:
The subject exempt plant change installed new concrete piers in support of the replacement of the Unit 1 Unit Auxiliary Transformer (UAT).
New concrete piers were required for the fire suppression deluge system which was redesigned due to physical differences between the existing GE and the new SMIT transformer.
Two other exempt changes were required to complete the replacement of the Unit 1 UAT:
E04-1-93-326 replaced the existing fire protection system piping and fire detection method.
The deluge piping was replaced due to the physical differences between the existing GE UAT and the new SMIT UAT.
The detection method was changed in order to make it more reliable.
The overall operation of the system did not change.
E04-1-93-327 reinstalled the transformer control circuitry.
These changes were necessary due to slight differences between the GE and SMIT transformers.
The control circuitry changes do not affect the operation of the plant.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or componer.t is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Loss of Auxiliary Power UFSAR SECTION 8.3.1 For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
TIrllOP3GAIITYW4MAY.RIT
i E04-1-93-325 CONTD 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the UAT is being replaced by a. newer transformer.
The failure mode of this new transformer, fire protection system, and control circuitry is the same as for the existing transformer.
The failure rate due to these changes is reduced due to the more reliable transformer and enhancements to the fire protection system.
Therefore, an accident different from those previously evaluated in the i
SAR is not created.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
t 1
l nmOMGMEWWGMY.MT
l l
PO4-1-91-127 Replacement of existing T-Quencher Bolts DESCRIPTION:
Installed new replacement welded and/or bolted threaded rod to replace rods anchoring the T Quencher supports located in the torus.
The sample of rods was removed for examination to confirm the absence of stress corrosion cracking.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
t 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because change does not affect equipment operations or functions.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because there is no change to Technical Specifications.
mCHOMGMEWWGMY.RW
oca 1o00-6 l
UNIT 1(2)
RE. VISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TR g
1._
Quad Cities Nucient Power Station Date: f N Reference No..a.er; 7 -. W
" R St-'*
flivre 0luun.- Onh t
9 Submitted by: bd AnaPo
+
n:
!FOR: REVIEW:gw%u M5 n +-
Safety Evaluations t[QIinvolving an unroviewed safety question as defined in 10CFR5 1.
for:
Changes to procedures as descdbed in the Safety Analysis Report.
a.
Changes to equipment or systems as described in the Safety Analysis Report.
b.
)
Tests or experiments.tiQI described in the Safety Analysis Report.
c.
Propceed changes which irwolve an unrevlowed safety question as defined in 10CFM50.5 2.
a.
Procedure changes.
b.
Equipmart or system changes.
Tase.7 eiuperiments.
c.
Proposed changes to the Technical Apaamia=*=is or Operating License.
3.
Noncompliance wth oodes, reguistions, orders, Technical Specifications, license 4.
requirements, or intamal f s_d=as or instructions having nucisar safety siv.
Signuicent operating abnormalties or deviations from normal and expected performa i
5.
plant equipment that asects nuclear solsty.
6.
AB REPORTAIKE EVENTS (LERs only).
AB recognhed indicultions of an unerticipated #4=Relancy in design or uf, A. of
".;y-l 7.
reisted structures, systems, or components. -
AR changes to the Station Emergency Plan prior to ime 8.
AB Roms referred by the Systems Engineering Supervisor. Station Manager, Ste Vice R
Prealdent, and Genomi Manager of Quesy Programs and Assessments.
e-
_,_ _ m. -- ~ w ~. m #..
.~i
!PDRInrurudaM--MbE9"&*Sf3*AWPW 4M@.W* -
l
[ 10. Other OSR leems/Documerts. tie addressed above.
This Transmatel is being made in accordance wth Qumd Chios Nuclear Power Station Technical SpecMcations s.1.G.2.d(1) for informselon only. No specific action is req unless doomed necessary by Oliske Review and Irwestigmove Function.
i 8
5 W
CGE QCAP 1100-9 UNIT 1(2) l REVISION O l
l ATTACHMENT G (Page 1 of 8) 10CFR50.59 SAFETY EVALUATION GENERAL INFORMATION:
Safety Evaluation Number:
SE 37 Document identifier: SAR (Software Activity Request) # 991 WedhebertTome apt. Wort aequest NN etc )
Unit (s): 1 & 2 l System (s): 207 (Rod Worth Minimizer)
Applicable Plant Mode (s): ALL em s e.t s.n.*v new swo,,
j Plant Mode Restriction (s): NONE List Multiple Procedures Affected Below:
Procedure Number -
l Procedure Nuinber I
^ Procedure Numtd Procedure Numbert.
~
QCoP 2o7-1 l
l CHANGE DESCRIPTION:
- 1. Describe the proposed change-The Rod Worth Minimizer (RWM) software will be upgraded to include two new features. First, a select block will be added to the current rod blocks. This will cause a rod block upon selection of an out-of-sequence j
control rod. This select block function willinclude an on/off 'toggie' on the RWM touch screen. The other i
change is the provision of an insert block signalin Rod Exercise mode as soon as a control rod is moved in one i
notch, in addition, a rod block will be applied to all other rods at this time. The blocks tre removed when the current rod is withdrawn to its original position. Currently the RWM applies a rod block only after a rod has traveled more than one notch past its target position.
i i
- 2. Reason for the change:
These changes are being made to address several control rod movement errors. The software changes are the corrective actions of NTS item #2650193007305.
- 3. Is the change:
X Permanent Temporary - Expected Duration:
i
QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 2 of 8) 10CFR50.59 SAFETY EVALUATION REFERENCE DOCUMENTS:
4.
List reference documents used which describe the structure, system, or component. Identify documents referenced even if no information was found in that section, s.
UFSAR Section(s): 7.7.2 (RWM),15.4.10 (Rod Drop Accident) b.
SER Section(s):
c.
Tech. Spec. Section(s): 3.3/4.3B (Control Rods) d.
Fire Protection Program Document Pkg Section(s):
e.
Code of Federal Regulations Section(s):
f.
Regulatory Guides /NUREGs:
g.
Other: RWM User's Documentation (Rev 3.1, October,1990)
EVALUATION:
5.
Describe how plant operation is affected when the structure, system, or component function is changed as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes, include a discussion of any changed interactions with other structures, systems, or components.
Using the new software, plant operation will be affected by the addition of the select block and the blocks applied in Rod Exercise mode. The select block is in addition to the current blocks and will be applied when an out-of-sequence rod is selected. This block willimprove plant operation in that it adds another barrier to the NSO selecting and moving an incorrect control rod. The Rod Exercise blocks currently take effect only after a rod has traveled more than one notch PAST its target position. The upgraded Rod Exercise blocks will take effect immediately at the target in position, in addition to blocking movement of all other rods while a rod is inserted. The Rod Exercise blocks will improve operation by adding a barrier to mispositionings during the rod exercise procedures.
No physical changes will be made to the RWM, so there will be no changed interactions with other structures, systems, or components after the software upgrade. All interlocks with the RWM will remain the same, as will all other existing rod blocks provided by the RWM.
6.
Describe how the change will affect equipment failures. Describe any new failure modes and their impact during all applicable operating modes.
3 The only changes that will be made are to the software of the RWM - no physical changes will be made. Allinteractions with other systems and components will be unchanged. As a result, the failure modes of the RWM will remain the same as before the software upgrade.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O i
ATTACHMENT G (Page 3 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system, or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
15.4.10 Rod Drop Accident 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting, or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be aHected. To determine factors affecting the specification, it is necessary to review the UFSAR and SER where the Technical Specification Bases section does not explicitly state the basis.
3.3.B.3 1
i i
9.
Will the change involve a Technical Specification revision?
YES X
NO If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing Step 14, indiocate that a Technical Specification revision is required.
i j
com QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 4 of 8) 10CFR50.59 SAFETY EVALUATION
)
EVALUATION (cont'd):
- 10. To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR may be increased, use one copy of these pages to answer the fo!!owing questions for each accident listed in Step
- 7. Provide rationale for all NO answers.
Affected accident: Rod Drop Accident UFSAR Section: 15.4.10
- a. May the probability of the accident be increased?
YES X
NO The operation of the RWM is independent of the Control Rod Drive (CRD) System and has no impact on the withdrawal of a rod. It provides rod blocks to the Reactor Manual Control System (RMCS), but does not affect the interaction of the control rods with the reactor internals. Because this interaction is the means by which a control rod becomes stuck and later drops to the full out position, the RWM cannot affect this event. As a result, the probability of the Rod Drop Accident will not be increased by this software upgrade.
i
- b. May the consequences of the accident (off-site dose) be increased?
YES X
NO The RWM impacts the Rod Drop Accident only by limiting the worth (reactivity) of control rods during a reactor startup by enforcing a withdrawal sequence. This sequence must follow Banked Position Withdrawal Sequence (BPWS) rules to 20% power, which limits the energy deposited in the fuel to 280 cal /gm. These rules are not affected by the software upgrade, nor is the method by which the RWM enforces them. As a result, the rod worths after the software upgrade will not be increased, in addition, the RWM does not provide any mitigating effects after the accident. Because the same BPWS rules wih be enforced (no increase in deposited onthalpy), and the RWM cannot mitigate the Rod Drop Accident after it occurs, the consequences of the accident will not be increased by the software upgrade.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 5 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- c. May the probability of a malfunction of equipment important to safety increase?
YES X
NO No changes will be made to the RWM computers to add the rod blocks in the Rod Exercise Mode and selection blocks for out-of-sequence rods. These are software modifications only. No physical modifications will be made to the RWM computers, and no interactions with other systems will be altered. The probability of malfunction will remain the same for the RWM.
- d. May the consequences of a malfunction of equipment important to safety increase?
YES X
NO The RWM is comprised of two independent computers designed to enforce the control rod withdrawal sequence. For the withdrawal of the first 12 control rods, one RWM must be operable. If one RWM computer failed, the other would be available. If neither were available for the first 12 rods, startup would not be permitted. This consequence will ternain the same after the new software is installed because startup will still be prohibited. After the first 12 rods are fully withdrawn, the RWM is required to be operable up to 20%
power. However, if both RWM computers have failed, an additional verifier may be used as a substitute.
Again, the consequences of this malfunction will not change with the new software, as a second verifier will still be required.
If any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
.=::::
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 6 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd)
- 11. Based on the answers to Questiens 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR?
YES X
NO Describe the rationale for your answer.
The changes that will be made are to the RWM software only. No modifications will be made to the RWM computers, nor will any interactions with other systems be changed. The blocks that are being added will serve as an additional barrier to control rod mispositionings and all of the current rod blocks will be retained.
Because no other systems will be affected, there will be no adverse system interactions or accidents created in other systems.
In addition, because the RWM itself will not be altered, the failure modes will remain the same as before the software upgrade. As a result, there will be no new type of RWM malfunction not evaluated in the UFSAR.
1 If any answer to Question 11 is YES, then an Unreviewed Safety Question exists, l
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 7 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists. Proceed to Step 14.
Technical Specification: 3.3.B.3 Determine which of the following is true for the above specification:
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. The actual acceptance limit need not be identified to determine that no reduction in margin of safety exists. Proceed to Step 13.
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s) or margin (s) below.
The applicable parameter or condition change is in a potentially non-conservative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request Nuclear Licensing assistance to identify the acceptance limit or margin for the Margin of Safety determination. List the limit (s) or margin (s) below.
X The change does not affect any parameters upon which the Technical Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to Step 14.
List Acceptance Limit (s)/ Margin (s) of Safety
- 13. Use the above limits identified in Step 12 to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination. Include a description of compensating factors used to reach that conclusion.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 8 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 14. Check one of the following:
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The Proposed change MUST NOT be implemented without NRC approval.
X No unreviewed Safety Question will result (Steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required. Indicate applicable type (s) below:
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
The change is a plant modification or minor plant change. Indicate applicable type (s) below:
i A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Licensing may authorize the installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.
Nuclear Licensing has authorized installation, but no operation, prior to receipt of the NRC approval of License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed S fet@uestion will result, and provides authority for installation only.
3 Preparer /Date:
~ [7 /h,#
Q-;a 4y
- 15. Documentation is adequate to support the above conclusion and the conclusion is valid.
[M,[
f-7-fY Reviewer /Date:
JS. Obtain a Safety Evaluation number from the Systems Engineering Clerk. Record on Page 1.
- 17. Leave 1 Safety Evaluation copy with Systems Engineering Clerk. File original with package.
Completed:
Systems Engineering Clerk initials: M(
Date: 5-9-7 (final)
A
V QCAP 1000-6 UNIT 1(2) j REVISION O
{
ATTACHMENT A (Page 1 Of 1) i OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station Reference Number:
j'E-N -()3J(
Date:
5- / 2 - 1'/
Sub)ect:
0COS
.2 3cv - l
{& v b GCOS A3cz - 5 1%- 6 Submitted by:
[w [d FOR REVIEW:
1.
Safety Evaluations NQIinvolving an unreviewed safety question as defined in 10CFR50.59 for:
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
c.
Tests or experiments NOT described in the Safety Analysis Report.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59.
2.
Procedure changes.
a.
b.
Equipment or system changes.
c.
Tests or experiments.
3.
Proposed changes to the Technical Specifications or Operating Ucense.
4.
Noncompliance with codes, regulations, orders, Technical Specifications, license requirements, or internal procedures or instructions having nuclear safety significance.
5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
All REPORTABLE EVENTS (LERs only).
7.
All recognized indications of an unanticipated deficiency in design or operation of safety-related structures, systems, or components.
8.
All changes to the Station Emergency Plan prior to implementation.
9.
All items referred by the Systems Engineering Supervisor, Station Manager, Site Vice President, and General Manager of Quality Programs and Asessments.
FOR INFORMATION:
- 10. Other OSR ltems/ Documents NQI addressed above.
This Transmittal is being made in accordance with Ouad Cities Nuclear Power Station Technical Specifications 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and investigative Function.
8
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 1 of 8) 10CFR50.59 SAFETY EVALUATION GENERAL INFORMATION:
Safety Evaluation Number:
SE-9'
-() 3 $
Document identifier:
OCOS 2300-1 Rev. 6 u
.u v. u.. w. a - %..,
Unit (s): I and 2 System (s): HPCI (2300)
Applicable Plant Mode (s): All Modes m.,s, e.,s e a.w.swo-Plant Mode Restriction (s): None List Multiple Procedures Affected Below:
Procedure Number Procedure Number Procedure Number Procedure Number OCOS 2300-5 Rev. 6 CHANGE DESCRIPTION:
- 1. Describe the proposed change:
This procedure is being revised to have the operator verify lube oil temperature indicating switch setpoints are set to their proper value as listed in the procedure. Also, the HPCI pump is verified to be filled and vented locally prior to rolling the HPCI turbine.
- 2. Reason for the change:
This procedure change is being performed to ensure proper setpoints prior to manual startup for routine surveillances to reduce the probability of spur'ious high temperature alarms which would require system L
shutdown.
l 3. Is the change:
X Permanent Temporary - Expected Duration:
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 2 of 8) 10CFR50.59 SAFETY EVALUATION REFERENCE DOCUMENTS:
4.
List reference documents used which describe the structure, system, or component. Identify documents referenced even if no information was found in that section.
a.
UFSAR Section(s): 6.3, 7.3, 9.3.3.9. 15.5.1, 15.6.2,15.6.5 b.
SER Section(s):
3.5.2.1 c.
Tech. Spec. Section(s):
3.5.C/4.5.C. 3.5.G/4.5.G, 3.7.A/4.7.A d.
Fire Protection Program Document Pkg Section(s):
e.
Code of Federal Regulations Section(s):
f.
Regulatory Guides /NUREGs:
O 1 ther:
EVALUATION:
5.
Describe how plant operation is affected when the structure, system, or component function is changed as intended (i.e., focus on system operation / interactions in the absence of equipment failures).
Consider all applicable operating modes. Include a discussion of any changed interactions with other structures, systems, or components.
These temperature indicating switches give alarms only to the Control Room Operator. They do not provide any trip functions. Verifying these setpoints will eliminate unnecessary alarms but yet will stil!
provide alarm protection as designed.
l 6.
Describe how the change will affect equipment failures. Describe any new failure modes and their impact during all applicable operating modes.
This change does not affect the operation of any equipment and therefore will not affect any failure modes By ensuring proper alarm setpoints, equipment damage can be averted due to operator response to high temperature conditions.
l
CGE QCAP 1100-9 UNIT 1(2)
)
REVISION O ATTACHMENT G (Page 3 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
i 7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system, or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
Accident UFSAR Section Small break LOCA 15.6.4, 15.6.5 i
8.
Ust each Technical Specification (Safety Umit Umiting Safety System Setting, or Umiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine factors affecting the specification, it is necessary to review the UFSAR and SER where the Technical Specification Bases section does not explicitly state the basis.
None N/A l
9.
Will the change involve a Technical Specification revision?
YES X
NO If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing Step 14, indiocate that a Technical Specification revision is required.
CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 4 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 10. To determine if the probability or the consequences of an accident or malfunction of equipment important to safey previously evaluated in the UFSAR may be increased, use one copy of these pages to answer the following questions for each accident listed in Step 7. Provide rationale for all NO answers.
Affected accident:
None UFSAR Section:
N/A
- a. May the probability of the accident be increased?
YES X
NO The probability of an accident will not be increased because this procedure change does not affect any equipment which is currently considered an initiator of any analyzed accident.
l
- b. May the consequences of the accident (off-site dose) be increased?
YES X
NO The consequences of an accident (off-site dose) will not be increased because automatic operation of HPCI is not affected by the temperature switches involved in this procedure change, thus, plant response to accidents is unaffected from previous analyses.
-- 1
CGE QCAP 1100-9
~
UNIT 1(2)
REV7310N O ATTACHMENT G (Page 5 of 8) 10CFR50.59 SAFETY EVALUATION i
EVALUATION (cont'd):
- c. May the probability of a malfunction of equipment important to safety increase?
YES X
NO The probability of a malfunction of equipment important to safety will not increase because by verifying the switch setpoints prior to rolling the HPCI turbine will only ensure the HPCI lube oil system will operate as designed.
- d. May the consequences of a malfunction of equipment important to safety increase?
YES
{
X NO The consequences of a malfunction important to safety will not increase because automatic operation of HPCI is not affected by the temperature switches involved in this procedure change, thus, plant response to accidents is unaffected from previous analyses.
t if any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 6 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
i
- 11. Based on the answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR?
YES X
NO Describe the rationale for your answer.
Verifying lube oil temperature indicating switch setpoints prior to rolling the HPCI turbine doesn't adversely impact systems or functions so as to create an accident of a type different from those previously evaluated in the UFSAR. This procedure change will only aid in preventing possible damage to the HPCI turbine and thus decreasing the probability of any accident.
r if any answer to Question 11 is YES, then an Unreviewed Safety Question exists.
CGE QCAP 1100-9 UNIT 1(2)
~
REVISION O ATTACHMENT G (Page 7 of 8) 10CFR50.59 SAFETY EVALUATION I
EVALUATION (cont'd):
i
- 12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists. Proceed to Step 14.
Technical Specification:
Determine which of the following is true for the above specification:
i All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. The actual acceptance limit need not be identified to determine that no reduction in margin of safety exists. Proceed to Step 13.
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. Ust the limit (s) or margin (s) below.
The applicable parameter or condition change is in a potentially non-conservatiis direction and the Technical Specification neither provides an acceptance limit ny explicitly references a limit in the UFSAR. Request Nuclear Ucensing assistance to identify the acceptancu limit or margin for the Margin of Safety determination. Ust the limit (s) or margin (s) below.
The change does not affect any parameters upon which the Technical Specivations are based; therefore, there is no reduction in the margin of safety. Proceed to Step 14.
j Ust Acceptance Umit(s)/ Margin (s) of Safety
- 13. Use the above limits identified in Step 12 to determine if the margb of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination. Include a description of I
compensating factors used to reach that conclusion.
i
\\
i
t CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 8 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 14. Check one of the following:
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The Proposed change MUST NOT be implemented without NRC approval.
X No unreviewed Safety Question will result (Steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a Ucense Amendment. Notify Station Regulatory Assurance and Nuclear Ucensing that a Technical Specification revision is required. Indicate applicable type (s) below:
The change is not a plant modification or minor plant change and will not be imp!emented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
The change is a plant modification or minor plant change. Indicate applicable type (s) below:
A revision to an existing Technical Specification is required. The change l
MUST NOT be installed until receipt of an approved Technical Specification revision.
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Ucensing may authorize the installation, but not operation, prior to receipt of NRC approval of the Ucense Amendment. If such authorization is granted, l
the block below should be checked.
Nuclear Licensing has authorized installation, but no operation, prior to receipt of the NRC approval of Ucense Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result, and provides authority for installation only.
Preparer /Date, hhd P/.2-@/
- 15. Documentation is adequate to support the above conclusion and the conclusion is valid.
Reviewer /Date:
h
- 762, f[2,[P/
- 16. Obtain a Safety Evaluation number from the Systems Engineering Clerk. Record on Page 1.
- 17. Leave 1 Safety Evaluation copy with Systems Engineering Clerk. F!ie original with package.
Completed:
Systems Engineering Clerk initials:
M Date: c;. Q.qf (final)
Q~.AP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMIT Quad Cities Nuclear Power Station f Reference Nuirit.,st:
k -C39 Date: 5-17-94 l
Subject:
one time extension to 67 dav Safe Shut Down Administrative Technical Recuirement-(ATR) for Unit 2 durinc 01R13.
f Subm!tted by:
- Jim Masterlark J
i n.n gv _. y n vv t
TOR ~REYlEW:fde b "
Safety EvaluationsRE involvmg an unrewwwed safety question as defined in 10CFR50.59 1.
for:
Changes to procedures as described in the Safety Analysis Report.
a.
Changes to equiprnent or systems as described in the Safety Analysis Report.
l b.
Tests or Wnts RQI described in the Safety Analysis Report.
c.
Proposed changes which involve an unroviewed safety question as defined in 10CFR50.59.
l 2.
a.
Procedure changes.
b.
Equipment or system changen.
c.
Tests or N 3.
Proaa-i changes to the Technical Specifications or Operating License.
i d
4.
Nors.plance wth codes, regulations, orders, Technical Specifiestions, license reautrernents, or intamal procedures or instructions having nuclear safety significance.
S;wi.lTet operating abnormalties or deviations from normal and expected pedormance of 5.
piart W
,,e.
that allocas nucdear sofory.
6.
AB REPORTABLE EVENTS 0.ERs only).
7.
AB i.csy.J d indications of an unsreerdp=hari deficiency in design or operation of safety-i
- related structures, systems, or components.
a.
All changes to the Stadon Lr.,v.,si lan prior to implementation.
P AB ltems referred by the Systems Engineering Supervisor, Station Manager, Ste Vice c
9.
Presidertt, and General Manager of Qualty Programs and Assessments.
__._,--mm _
IFOR:1NFORMATIONMM@@%x.,s.WewYeN#%w -
1
- 10. Other OSR ltems/DocumernsRQI andressed above.
This Transmittal is being made in accordance with Quad Cities Nuclear Power Station Tm).h Sps;Te. 6.1.C.2.d(1) for information only. No specific action is required unless deemed necessary by Offste Review and investigative Function.
e
CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 1 of 8) 10CFR50.59 SAFETY EVALUATION GENERAL INFORMATION:
Safety Evaluation Number:
SE-74 039 Document identifier: 67 Day SSD ATR Extension for SSD Path B v,ese. Tem a m eene non. m Unit (s): 1 and 2 System (s): 4100, 287, 1000, 1300, 2900, 6600, 6700, 7300,8300,8350 l
l Applicable Plant Mode (s): All modes monswun+wsweew nes.,swe m Plant Mode Restriction (s): No restnchons Ust Multiple Procedures Affected Below:
Procedure Number.
Procedure Nun:bar.
Procedure Number Procedure Number l
l CHANGE DESCRIPTION:
[
- 1. Describe the proposed change:
This change to the fire protection program will have two parts:
-allow the extension of the 67 day safe shutdown ATR for safe shutdown path B by 24 days
-establish additional compensatory measures.
(See Attached document)
- 2. Reason for the change:
This change wiH allow the work to continue without requiring the operating unit to shutdown wtan the 67 day ATR expires. Additional compensatory measures will be established to ensure safe operation during the extension of the ATR.
(See Attached document)
- 3. Is the change:
Permanent X
Temporary - Expected Duration May 21 thru June 15
CGE QCAP 1100-9 j
UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 2 of 8) 10CFR50.59 SAFETY EVALUATION REFERENCE DOCUMENTS:
4.
List reference documents used which describe the structure, system, or component. Identify documents referenced even if no information was found in that section.
a.
UFSAR Section(s): 9.5.1 b.
SER Section(s): Fire Protection Reports, Volume 3-Fire Protection SERs c.
Tech. Spec. Section(s): 6.0 Administrative Requirements d.
Fire Protection Program Document Pkg Section(s): Fire Protection Reports, Volume 2, Safe Shutdown Reports, Section Administrative Technical Requirements e.
Code of Federal Regulations Section(s): 10CFR50 Appendix R,10CFR50.48 f.
Regulatory Guides /NUREGs: None g.
Other: GL 86-10, GL 88-12, Attached Documentation (Justification)
EVALUATION:
5.
Describe how plant operation is affected when the structure, system, or component function is changed as intended (i.e., focus on system operation / interactions in the absence of equipment failures).
Consider all applicable operating modes. Include a discussion of any changed interactions with other structures, systems, or components.
This is an administrative change to the Fire Protection Program. This change impacts the compensatory measures required when Safe Shutdown Eqttipment is Inoperable. This change will allow equipment to be inoperable for more than 67 days if compensatory measures are estblished which would compensate the for the weekness in the 3rd echelon for fire protection (See Step 10.a).
6.
Describe how the change will affect equipment failures. Describe any new failure modes and their impact during all applicable operating modes.
There are no new failure modes. The plant would continue to operate within the bounds of the ATR. The only change is that the ATR would be extended and additional compensatory measures will be estabitshed.
1 l
CGE QCAP 1100-9 UNIT 1 (2)
REVISION 0 i
ATTACHMENT G (Page 3 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system, or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
Appendix R Fire as described in the Fire Hazards Analysis i
8.
List each Technical Specification (Safety Umit, Umiting Safety System Setting, or Limiting Condition for i
Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine factors affecting the specification, it is necessary to review the'UFSAR and i
SER where the Technical Specification Bases section does not explicitly state the basis.
)
None 9.
Will the change invohre a Technical Specification revision?
YES X
NO If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When mmpleting Step 14, indicate that a Technical Specification revision is required.
i
CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 4 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 10. To determine if the probability or the consequences of an accident or malfunction of equipment impodant to safety previously evaluated in the UFSAR may be increased, use one copy of these pages to answer the following questions for each accident listed in Step 7. Provide rationale for all NO answers.
Affected accident: Appendix R Fire
- a. May the probability of the accident be increased?
YES X
NO Basis of Fire Protection Program:
The fire protection program is based upon the concept of defense in depth. This defense consists of three echelons of protection between fire initiators and a possible uncontrolled release. No one of these echelons is perfect or complete by itself. These echelons are as follows:
- 1. FIRE PROTECTION PROGRAMS TO PREVENT FIRE INITIATION: This echelon helps to ensure that a fire does not initiate, or if R does initiate, that it does not expand beyond the incipient stages. These programs include the transient combustible programs, welding and gnnding permits, and good housekeeping practices.
- 2. RAPID SUPPRESSION AND DETECTION: This eenelons win help to ensure that a fire that is in Hs incipient stage does not propagate into an
- Appendix R" fire where safe shutdown of a unit will be required. This barrier consists of numerous automatic detection and suppression systems located throughout the power block.
- 3. FIRE BARRIERS / SAFE SHUTDOWN: This barrier consists of two parts, the first part is physical fire barners that separate safe shutdown equipment from a Design Basis Fire as described in the Fire Hazards Analysis. The second part is the safe shutdown equipment' procedures that would safely shut down a unit while the fire is contained.
If an Appendix R fire (as described in the Fire Hazards Analysis) were to occur when a safe shutdown path is inoperable, safe shutdown could not be achieved within the bounds of the safe shutdown procedures or within the required time limits meet the requirements by Appendix R. Therefore, compensatory measures and S7 day ATR are established to help ensure that a fire will not reach the Appendix R fire stage.
Basis for ATR extension:
When the 67 day ATR is exceeded on 5/21/94, additional compensatory measures will be initiated to further reduce the probability that a fire could reach the Appendix R Fire stage where safe shutdown would be required These compensatory measures described within this document win strengthen the first two bamers to make-up for d reduction in the third. These compensatory measures will be limited to 24 days. This wiU allow work to continue on Unit 1 systems during fra refueling outage. After expiration of the 24 days, if all safe shutdown path B is not retumed to operabuity, Unit 2 will be required to shutdown.
The extension of the ATR in itself willincrease the probability of an Appendix R fire. However, tne increase in compensatory measures will decrease the probability of a fire occurring and wiD increase the probability of mitGatir g a small fire br. fore it becomes an Appendix R fire. Therefore, the overall probability for the accident to occur is equal or less than WPnout this extension.
- b. May the consequences of the accident (off-site dose) be increased?
YES X
NO The change to the, program allows 24 additonal days to the safe shutdown ATR whHe safe shutdown path B wiu not be avauable. There will be no change to the consequences of the Appendix R hre when compared to the original ATR criteria.
In addition, compensatory measure will be established to mitigate the consequenses of an incipiant fire.
I i
CGL QCAP 1100-9 UNIT 1 (2) l REVISION 0 I
ATTACHMENT G (Page 5 of 8) i 10CFR50.59 SAFETY EVALUATION f
EVALUATION (cont'd):
- c. May the probability of a malfunction of equipment important to safety increase?
YES X
NO i
i This change is administrative in nature and does not affect plant equipment.
Therefore, this change does not affect the probability of maNunction. Compensatory measures will help to ensure that a fire does not occur to i
challenge plant equipment.
- d. May the consequences of a malfunction of equipment important to safety increase?
i YES X
NO l
This is an administrative change and does not affect the operation of equipment. This change allows safe j
shutdown equipment to be inoperable while compensatory measures are established to reduct the probability of a design basis fire to occur.
j i
i l
l i
If any answer to Question 10 is YES, then an Unreviewed Safety Question exi.sts.
i l
I i
e
~ _ - ----..
CGE QCAP 1100-9 UNIT 1 (2)
REVISION 0 ATTACHMENT G (Page 6 of 8) j 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
i
- 11. Based on the answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibihty of an accident or malfunction of a type different from those evaluated in the UFSAR?
YES X
NO Describe the rationale for your answer.
l This is an administrative change to the procedure and therefore does not effect equipment operation. The type of accident that could occur (Appendix R Fire) has already been evaluated. The stations approved program addresses the consequences of this fire. No new accident types will be created by this change.
I If any answer to Question 11 is YES, then an Unreviewed Safety Question exists.
I t
i i
i
~.
l C G r.,
QCAP 1100-9 UNIT 1 (2)
REVISION 0 l
ATTACHMENT G (Page 7 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists. Proceed to Step 14.
Technical Specification:
l Determine which of the following is true for the above specification:
l All changes to the parameters or conditions used to establish the Technical i
Specification requirements are in a conservative direction. The actual acceptance limit need not be
)
identified to determine that no reduction in margin of safety exists. Proceed to Step 13.
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. Ust the limit (s) or margin (s) below.
The applicable parameter or condition change is in a potentially non-conservative
[
direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request Nuclear Ucensing assistance to identify the acceptance limit or margin for the Margin of Safety determination. Ust the limit (s) or margin (s) below.
X The change does not affect any parameters upon which the Technical Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to Step 14.
List Acceptance Umit(s)/ Margin (s) of Safety i
i>
L i
- 13. Use the above limits identified in Step 12 to determine if the margin of safety is reduced (i.e., the new values l
exceed the acceptance limits). Describe the rationale for your determination..nclude a description of compensating factors used to reach that conclusion.
l l
i I
CGE QCAP 1100-9 UNIT 1 (2)
REVISION 0 ATTACHMENT G (Page 8 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 14. Check one of the following:
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The Proposed change MUST NOT be implemented without NRC approval.
X No unreviewed Safety Question will result (Steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a Ucense Amendment. Notify Station Regulatory Assurance and Nuclear Ucensing that a Technical Specification revision is required. Indicate applicable type (s) below:
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
The change is a plant modification or minor plant change. Indicate applicable type (s) below:
A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Ucensing may authorize the installation, but not operation, prior to receipt of NRC approval of the Ucense Amendment. If such authorization is granted, the block below should be checked.
Nuclear Ucensing has authorized installation, but no operation, prior to receipt of the NRC approval of Ucense Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result, and provides_ authority for installation only.
Preparer /Date:
((
k h [h y
5[a[g
- 15. Documentatio[is adequate to support the above conclusion and the conclusidn is valid.
Reviewer /Date:
[
[
.h-[fh l
- 16. Obtain a Safety Evaluation number from the Systems Engineering Clerk. Record on Page 1.
- 17. Leave 1 Safety Evaluation copy with Systems Engineering Clerk. File original with package.
Systems Engineering Clerk initials:
htC Da s:
c5 - 1 % -.
(final)
u QCAP 1000-6 UNIT 1(2)
REVZSZON'O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station
(..
Reference Number:
'14 - o42.
Date: 5/23/7 4
Subject:
T~ m o. A H. 94 3 2.
DissdlLa 9ersud Dlor.1 aw e
)
ue 2.
Submitted by:
M o Siam <3 I
1 Ii 6in i i % 5 W % 5 6 ll 7 1.
Safety Evaluations.!KII involving an unreviewed safety question as defined in 10CFR50.59 for:
~
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equipment or systems as described in the Safety Ansiysis Report.
c.
Tests or experiments.tiQI described in the Safety Analysis Report.
2.
Proposed changes which involve an unroviewed safety question as defined in 10CFR50.59.
a.
Procedure changes.
b.
or @ Q,,
c.
Tests or superiments.
3.
Proposed changes to the Technical SpecEications or Operating Uconse.
4.
Noncompliance wth codes, regulations, orders, Technical R&, license i
requirements, or intamal procedures or instructions having nucisar safety signricance.
s.
signuicent openning abnormenties or deviations from nommi and expected performance of plant equipment that asects nucisar safety.
1 6.
M REPORTABl.E EVENTS 6.ERs only).
1 7.
M recogniend indicadons of an unernicipseed danciency in design or operation of safory-reisted structures. systems, or components. -
J a.
M changes to the Station Emergency Plan prior to implementation.
4
- s. M ltems referred by the Systems Engineering Supervisor, Station Manager, Site Vice Presidert, and General Manager of Ousuty Programs and Assessments.
fhiENMM}NMbdMNA5Md$"MN-
- 10. Other OSR leems/ Documents.tElIaddressed stuws.
This Transmktal is beira made in accordance with Quad Cities Nuclear Power Station Technical W 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary try Offsite Review and Irt. - ^(-% Function.
8
CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 1 of 8) 10CFR50.59 SAFETY EVALUATION idENEN CINhdNMATidND#dMA$ne$$Kidd!inilseeF$$$5d$416 '<NWR ik5@MMin$$$$Bsy Safety Evaluation Number:
SE-94 C4 2.
Document identifier:
Temp.All. % - 2.- 3 2.
(Modecaton, Temp. AL. Work Request Number, etc )
Unit (s): Two System (s): 010 Applicable Plant Mode (s): All modes (Run, MartuoMat Standby, Refuel, SN)
Plant Mode Restriction (s): No restrictions List Multiple Procedures Affected Below-dNEAESudNumber
$dM[Procedur$ NudseI*5 d MM[N$$$ dure Number
'edNEiLiM E M ) JW x
NONE
[dNANGE[DESCRilNjbN34^ NNNd W
E@ ' O '* @@ '$NI?in [ M
@ '
'/^',, i ~
,s v
./'l',
'.h,,
- 11. Based on the answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR?
YES X
NO Describe the rationale for your answer.
This following changes have been anade to the Unit 2 PersonnelInterlock Doors for Primary Containment _
The interlocks that prevent rnore than one door open at a time have been defeated and are being controlled as per Technical e
specifications 3J.AJ.b and c.
One interlock door closed will ensure the integrity of Primary Containtnent Administrative controls, as required by Ts, are being implemented that will ensure only one door is open at a time.
i A valve that is connected to the interlock mechanism that equalizes pressure between the Drywell and volume between the air lock doors has been gagged in the CLOSED Position.
The air lock doors are being leak rate tested to verify that there is no leakage or acceptable leakage out of the air locks. The valve that communicates the Drywell volume and air lock volume has been gagged in the CLOSED postbon to ensure that during a seismic event and / or a decrease in reactor coolant that the vatve will remain in the CLOSED (as tesied) condition.
The strongbacks" have been left installed.
a The strongbacks have been installed and will be left on the Drywell air lock door. The strong back is a series of structural steel that bolts onto the Drywell door and secures the door in the closed postbon for leak rate testing. The strongbacks have been previously evaluated and found acceptable for operation.
Based on the above information and the fact that the Technical Specirication LCO is being implemented the possibility of an accident or malfuncbon of a type d:fferent from those evaluated in the UFEAR is not created.
I If any answer to Question 11 is YEE, then an Unreviewed Safety Question exists.
I
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 7 of 8) 10CFR50.59 SAFETY EVALUATION ENNLU TE05(hohd)sili%heJ#$f#M56 4%s $f$$@ Eg
/
- .i ["$ M ~ C J4 / y[.- #
- 12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists. Proceed to Step 14.
Technical Specification:
Determine which of the following is true for the above specification:
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. The actual acceptance limit need not be identified to determine that no reduction in margin of safety exists. Proceed to Step 13.
X The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. Ust the limit (s) or margin (s) below.
The applicable parameter or condition change is in a potentially non-conseivative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request Nuclear Ucensing assistance to identify the acceptance limit or margin for the Margin of Safety determination. Ust the limit (s) or margin (s) below.
The change does not affect any parameters upon which the Technical Specifications l are based; therefore, there is no reduction in the margin of safety. Proceed to Step 14.
Ust Acceptance Umit(s)/ Margin (s) of Safety 3.7.A.7.b / 3.7.A.7.c
- 13. Use the above limits identified in Step 12 to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination. Include a description of compensating factors used to reach that conclusion.
The LCO requirernents for an air lock door inoperable and the air lock interlock mechaniern inoperable. Therefore, the Technical Specifications will be meet. This will ensure the rnargin to safety will be rnalntained.
CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 8 of 8) 10CFR50.59 SAFETY EVALUATION
[EUA(ONTION (cont'd): 93 *!D/ ['W'f:Wt,'bi" r 'N? M'E$?"N^ 'h^ ^ f' :' ":WN M D:n 7;
- 14. Check one of the following:
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The Proposed change MUST NOT be implemented without NRC approval.
X No unreviewed Safety Question will result (Steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a Ucense Amendment. Notify Station Regulatory Assurance and Nuclear Ucensing that a Technical Specification revision is required. Indicate applicable type (s) below:
l The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
The change is a plant modification or minor plant change. Indicate applicable type (s) below:
A revision to an existing Technical Specification is required. The change l
MUST NOT be installed until receipt of an approved Technical Specification revision.
l The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. in these cases, Nuclear Ucensing may authorize the installation, but not operation, prior to receipt of NRC approval of the Ucense Amendment. If such authorization is granted, i
the block below should be checked.
l 8
Nuclear Ucensing has authorized installation, but no operation, prior to receipt of the NRC approval of Ucense Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result, and provides authority for instaliation only.
Preparer /Date: Ot h 5 /u r,s
- 15. Documentation is adequate to support the above conclusion and the conclusion is valid.
Reviewer /Date:[M[
/[
[-f.,2-f[
i i
- 16. Obtain a Safety Evaluation number from the Systems Engineering Clerk. Record on Page 1.
- 17. Leave 1 Safety Evaluation copy with Systems Engineering Clerk. File original with package.
Completed:
Systems Engineering Clerk initials:
}y Date:
5_ p t;. qq l
QCAP 1000-6 UNIT 1(2) arvIszow o ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL g
Quad Cities Nuclear Power Station t..
j Reference Number: 5E-74-o43-Date: F/29/94 Subject SE.s8 # 4-2)f6 Submitted by: u/ / 42- / M1 1,i9 M 6
.a n TOR:REVIEWh.~.-. g @,,dW!o 1.
Safety EvaluationsEQIinvolving an unroviewed safety question as defined in 10CFR50.59 for:
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
c.
Tests or experiments RQI described in the Safety Analysis Report.
2.
Proposed changes which involve an unreviewed safety question as denned in 10CFR50.59.
Procedure changes.
a.
~
b.
Equipment or system changes.
c.
Tests or esperiments.
3.
Proposed changes to the Technical ap esir ein,is or Operating 1.Jconse.
4.
Noncompuence wth codes, reguistions, orders, Technioni Specmastions, license requirements, or intamal procedures or instructions having nuclear safety significance.
S.
Signmcant operating abnormouties or deviations from normal and expected performance of plant egulpmart that aAsces nuclear safety.
a.
M ROORTABE EVENTS 6.ERs only).
7.
AB rb.Agnized intScutions of an unanticipated deficiency in design or operation of safety-reisted mrucases, erstems, or componsens. -
a.
AR chenpas to the Station Emergency Plan prior to implemanation.
- s. m tems retened by the Symems Engineering Supervisor, motion Manager, Site Vice President, and General Manager of Quatty Programs and Assessments.
ib TR 5 5 Uh?bIbb5 5 $ 5hiM9 5 !
- 10. Other OSR lisms/DonasnentsEE addressed above.
This Transmutal is being made in accordance wth Oumd Chios Nuclear Power Station Technical AW a.1.G.2.d(1) 9er information only. No specific action is required unless deemed necessary by Offsite Review and Irwestigsdve Function.
f 8
CGE QCAP 1100-9 UNIT 1(2) l REVISION O ATTACHMENT G (Page 1 of 8) 10CFR50.59 SAFETY EVALUATION GENERAL INFORMATION:
Safety Evaluation Number:
SE-hh Ch3 Document identifier: Work Request 011619 & Q11620, SESR #4-2156 DA edd cedon. Temp AM. Wort Recussi Nurnber. etc )
l System (s): 1100 Unit (s): 1 Applicable Plant Mode (s): All modes tRst 9'artmMot 9'auftev. w# 'JPudown)
Plant Mode Restriction (s): No restrictions List Multiple Procedures Affected Below:
Procedure Number Procedure Number Procedure Number Procedure Number None CHANGE DESCRIPTION:
- 1. Describe the proposed change:
The installation of a larger U-bolt (5/8" versus 1/2") on pipe support M-987D-75 because of increased weight of a parts upgrade of the Standby Liquid Control (SBLC) Accumulators for Unit 1. The replacement accumulators evaluation is ME-93-0541-00, Revision 2.
- 2. Reason for the change:
The original accumulators are no longer available from the manufacturer and the replacement accumulator weighs 95 lbs. which is more than 67 lbs., the weight of the original accumulator. This increase in weight has been evaluated (SESR 4-2156) and requires a larger U-bolt (5/8" versus 1/2") be insta!!ed on pipe support M-987D-75 to support design loads.
- 3. Is the change:
X Permanent Temporary - Expected Duration:
1 i
CGE QCAP 1100-9 t
l UNIT 1(2)
REVISION O ATTACHMENT G (Page 2 of 8) 10CFR50.59 SAFETY EVALUATION REFERENCE DOCUMENTS:
4.
List reference documents used which describe the structure, system, or component. Identify documents referenced even if no information was found in that section.
a.
UFSAR Section(s): 3.0, 3.2, 3.7, 3.9, 4.6, 9.3.5 b.
SER Section(s): None c.
Tech. Spec. Section(s): 3.4/4.4 d.
Fire Protection Program Document Pkg Section(s): None e.
Code of Federal Regulations Section(s): None f.
Regulatory Guides /NUREGs: None g.
Other: DBD-OC-139, Rev A; Vetip manual C0116; Drawing C68514-200, Drawing M-40, Rev AG EVALUATION:
5.
Describe how plant operation is affected when the structure, system, or component function is changed as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes. Include a discussion of any changed interactions with other structures, systems, or components.
The SBLC operates in the same manner as prior to installation of the heavier accumulator and larger U-bolt on support M-987D-75. The larger U-bolt performs the same functions as the smaller U-bolt. The function is to provide support in a seismic event and restrain pipe movement. The interactions with other structures, systems and components does nogchange.
9&mrut*M i/*
6.
Describe how the change will affect equipment failures. Describe any new failure modes and their impact during all applicable operating modes.
The larger U-bolt will not affect any equipment failures that had not been considered in earlier evaluations. There are no new failure modes which would occur from the insta!!ation of a larger U-bolt.
The larger U bolt performs the same design function as the U-bolt being replaced. A larger U-bolt is required due to the increased weight of the new accumulator.
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CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 3 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
e The changed structure, system, or component is explicitly or implicitly assumed to function during or after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
Anticipated Transient Without SCRAM UFSAR Section 9.3.5
( includes seismic) 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting, or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine factors affecting the specification, it is necessary to review the UFSAR and SER where the Technical Specification Bases section does not explicitly state the basis.
Standby Liquid Control System Technical Specification 3.4/4.4 9.
Will the change involve a Technical Specification revision?
YES X
NO If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing Step 14, indiocate that a Technical Specification revision is required.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O i
ATTACHMENT G (Page 4 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 10. To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR may be increased, use one copy of these pages to answer the following questions for each accident listed in Step
- 7. Provide rationale for all NO answers.
Affected accident: Ant!cipated Transient Without UFSAR Section: 3.9 Scram (ATWS) (includes seismic)
- s. May the probability of the accident be increased?
YES X
NO The larger U-bolt and increased weight from the replacement accumulators do not increase the probability of the ATWS accident. The SBLC system is required in the event that an accident would occur.
- b. May the consequences of the accident (off-site dose) be increased?
YES X
NO The consequences of the accident are not increased. The SBLC system will perform its design function with the increased weight and installation of the larger U-bolt. This is based upon SESR 4-2156, which evaluated the increased weight of the replacement accumulator.
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CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 5 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- c. May the probability of a malfunction of equipment important to safety increase?
YES X
NO i
The probability of a malfunction of equipment important to safety does not change. The U-bolt function does not change and therefore does not change the design basis function of the SBLC system. SESR 4-2156 Ovaluated the design loading for SBLC piping and determined that with the larger U-bolt, the SBLC will meet loading requirements. Therefore, the probability of a malfunction of equipment does not increase.
- d. May the consequences of a malfunction of equipment important to safety increase?
YES X
NO l
The consequences of a malfunction of the U-bolt remain the same. The function of the larger U-bolt is the I
same function as the smaller U-bolt. The larger U-bolt operates in the same manner as the smaller U-bolt.
t l
If any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
]
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CGE QCAP 1100-9 UNIT 1(2)
REVISION 0 ATTACHMENT G (Page 6 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 11. Based on the answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR?
YES x
NO Describe the rationale for your answer.
The larger U-bolt does not create the possibility of an accident or malfunction of a type different from those ovaluated in the USFAR. The larger U-bolt functions in the same manner as the smaller U-bolt it is replacing.
The higher weight of replacement SLC accumulators requires the larger U-bolt to maintain the system seismically. Sizing of the U-bolt has been evaluated by seismic calculation (SESR 4-2156).
If any answer to Question 11 is YES, then en Unreviewed Safety Question exists.
i i
i
CGE QCAP 1100-9 UNIT 1(2) i REVISION O ATTACHMENT G (Page 7 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists. Proceed to Step 14.
Technical Specification: Standby Liquid Control System,3.4/4.4 Determine which of the following is true for the above specification:
X All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. The actual acceptance limit need not be identified to determine that no reduction in margin of safety exists. Proceed to Step 13.
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s) or margin (s) below.
The applicable parameter or condition change is in a potentially non-conservative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request Nuclear Licensing assistance to identify the acceptance limit or margin for the Margin of Safety determination. List the limit (s) or margin (s) below.
The change does not affect any parameters upon which the Technical Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to Step 14.
List Acceptance Limit (s)/ Margin (s) of Safety
- 13. Use the above limits identified in Step 12 to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination, include a description of compensating factors used to reach that conclusion.
SESR 4-2156 evaluated the increase in weight of the replacement accumulator and the replacement of a larger U-bolt on pipe support M-987D-75. The evaluation determined that these changes were within the design loadings of the SBLC system.
CGE QCAP 1100-9 UNIT 1(2)
REVISION O ATTACHMENT G (Page 8 of 8) 10CFR50.59 SAFETY EVALUATION EVALUATION (cont'd):
- 14. Check one of the following:
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The Proposed change MUST NOT be implemented without NRC approval.
X No unreviewed Safety Question wi!! result (Steps 10,11, and 13) AND no Techrical Specification revision will be involved. The change may be implemented in pecordance with applicable procedures.
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revd.;n is required. Indicate applicable type (s) below:
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
The change is a plant modification or minor plant change. Indicate applicable type (s) below:
A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. in these cases, Nuclear Licensing may authorize the installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.
Nuclear Licensing has authorized installation, but no operation, prior to receipt of the NRC approval of Lic9nse Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Ouption will result, and provides authority for installation only.
Preparer /Date:
ff 77
- 15. Documentatidb quate to support the above conclusion and the conclusion is valid.
Q A pWh()
g-/Z.f/9 Y Reviewer /Date:
- 16. Obtain a Skty Evaluation number from the Systems Engineering Clerk. Record on Page 1.
i
- 17. Leave 1 Safety Evaluation copy with Systems Engineering Clerk. File original with package.
Completed:
Systems Engineering Clerk initials:
M Date: $.ggr.(
(final)
I QCAP 1000-6 UNIT 1(2)
REVISION O l
ATTACHMENT A (Page 1 of 1) i OFFSITE REVIEW AND INVESTIGAT!VE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station Reference Number: M(4 90-003 Date: S/li /94 i
Subject:
CA0 REPH Roon A/c isruuanod I
Submitted by: kl[ll (g)(>f I
f I
I FOR REVIEW:
1.
Safety Evaluations.tiQIinvolving an unreviewed safety question as defined in 10CFR50.59 for:
j a.
Changes to procedures as described in the Saf?ty Analysis Report.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
c.
Tests or experiments.tiQI described in the Safety Analysis Report.
2.
Proposed changes which involve an unroviewed safety question as defined in 10CFR50.59.
a.
Procedure changes.
b.
Equipment or system changes.
c.
Tests or experiments i
4 i
3.
Proposed changes to the Technical Sp0cifications or Operating License 1
Nor.cuir.f ance with codes, regulations, orders, Technical Specifications, license li 4.
requirements, or intomal procedures or instructions having nuclear safety signifcance.
5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
M REPORTABLE EVEfRS (LERs only).
7.
M recognized indications of an unanticipated deficiency in design or operation of safety-related structures, systems, or components.
8.
M changes to the Station Emergency Plan prior to implementation.
9.
M ltems referred by the Systems Engineering Supervisor, Station Manager, Site Vice President, and General Manager of Quality Programs and Assessments.
- FOR"pfFORMATION *
[
- 10. Other OSR ltems/DocumentsRQIaddressed above This Transmittal is being made in accordance with Quad Cities Nuclear Power Station Technical Specifications 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and Investigsthe Function.
)
8
/
QAP 1100-521 Revision 1
(]
10CFRSO.59 SAFETY EVALUATIONS December 1990 V
Safety Evaluation Number:
SE-9/ - I4 /
Document Identifier:
8,4 fr- 0o3 (Modification, Temp Alt, Work Request Number, etc.)
Unit (s):
I System (s):
I"10#
Applicable Plant Mode (s):
4LL A *Jes Sply (RUN, STARTUP/hDT'STNBY, REFUEL or SHUTDOWN) 1.
Describe the proposed change:
fr om de. r a 6 Ls.u [.<
S C/0 Kess:< Keen.
No di Ldt will l-sLil
&n a:e e.. l. J ' e.. d a -s * < l.e&J /J s:Je.e L e ree~ a.J -- a:e L-Jtr..
a.:4 laed.J t-ssh 4Le r een. ft.da a I
><-ar w:ll fa n s us t ra ) CrJs.
CF m e e.
wL:eL w:t t r a n ist a 4La er: slid w s tin, L.a h'ee v2t-s-/.
e g
a J
2.
Reason for the change:
frovi da. e.. lone for ma: dana.ee oe unena l -ka n -,, hts.
e balr.d as a a w, s. TL:n "is c
t*e e 4,e ~ a.a
E n L. a o ~. 4 t* roar
':4.~.
3.
Is the change:
(*)
Permanent
(
) Temporary - Expected Duration:
Plant Mode (s) Restrictions:
4.
List the reference documents reviewed which describes the structure, system or component.
(Identify documents referenced even if no information was found in that section.)
a.
UFSAR Section(s):
/. J. 5
/s. /a. /
/J.J.J.i b.
SER Section(s):
/v, e.
c.
Tech Spec Section(s):
J. /a /S, /,. O v./.2
/
)
d.
Fire Protection Program Document Pkg Section(s): F#A f. f. y / 8, e. /ffa j
)
e.
Code of Federal Regulations Section(s):
Ne-c.
O f.
Regulatory Guides (NUREGs): JE d it,/i.
PF- /I.
J APPRoygg 1
4100a o.c.o.s.R.
m
0AP 1100-S21 Revision 1 I
5.
Describe how the change will affect plant operation when the changed structure, system or component function as intended (i.e., focus on 1
system operation / interactions in the absence of equipment failures).
Consider all applicable operating modes.
Include a discussion of any changed interactions with other structures, systems or components.
Sea. p //~s L <~f.
6.
Describe how the change will affect equipment failures.
In particular, describe any new failure modes and their impact during all applicable operating modes.
Se,_ ALL~a d.
7.
Identify each. accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true.
The change alters the initial conditions used in the UFSAR analysis The changed strur,ture, system or component is explicitly or O
=
-d*
implicitly assvLed to function during or after the accident Operation or structure, system or component failure of the changed structure, system or component could lead to the accident ACCIDENT UFSAR SECTION
_%. e.
8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected.
- Nose, r
r\\
. V, APPROVED l
DEC 311990 4100a Q.C.O.S.R.
P
(
GAP 1100-521 Revision 1 i
(
9.
Will the change involve a Technical Specification revision?
V)
(
) Yes
( X ) No If a Technical Specification revision is involved, the change cannot be i
implemented until the NRC issues a license amendment.
When completing step 14, indicate that a Technical Specification revision is required.
10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR may be increased, use one copy of this page to answer the following questions for each accident listed in step 7.
Provide the rationale for all N0 answers.
Affected accident N/A UFSAR Section:
May the probability of the accident be increased?
(
) Yes
(
) No May the consequences of the accident (off-site
(
) Yes
(
) No dose) be increased?
s
()
May the probability of a malfunction of equipment
(
) Yes
(
) No important to safety increase?
May the consecuences of a malfunction of equipment (
) Yes
(
) No important to safety increase?
If any answer to Question 10 is YES. then an Unreviewed Safety Question exists.
i APPROVED DEC 318 4100a C}.CLC).Si II-
)
I
QAP 1100-521 Revision 1 i
O, 11.
Based on your answers to Questions 5 and 6 does the change adversely tapact systems or functions so as to create the possibility of an et.cident or malfunction of a type different from those evaluated in the llFSAR?
(
) Yes
( X ) No l
Describe the rationale for your answer.
4* Q <' h~s 5 *~I l-0 :s euss o J s'~ r esf~n O
b i
If the answer to Question 11 is Yes. then an Unreviewed Safety Question exists.
I t
E APPROVED
,]
DEC 311990 4100a 4-Q.C.O.S.R.
i
l OAP 1100-521 Revision 1 n-Q 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following i
questions for each Technical Specification listed in step 8.
If no technical Specifications are impacted, then no reduction in margin of safety exists, proceed to stap 14.
f Technical Specification MA 1
Determine which of the following is true for the above specification:
(
)
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety l
exists, proceed to question 13.
(
)
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition.
List the limit (s)/ margin (s) below.
j
(
)
The applicable parameter or condition change is in a potentially non-conservative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR.
Request Nuclear Licensing assistance to identify the' acceptance limit / margin for the Margin of Safety determination.
List the O,
limit (s)/ margin (s) below.
List Acceptance Limit (s)/ Margin (s) of Safety r
13.
Use the above limits to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
l If a Marcin of Safety is reduced. an unreviewed Safety Question exists.
O
^eaovso DEC 311990 4100a Q.C.O.S.R.
A
I QAP 1100-521 Revision i 14.
Check one of the following:
(
) An Unreviewed Safety Question was identified in step 10, step 11, or step 13.
The proposed change MUST NOT be implemented without NRC approval.
( X)
No Unreviewed Safety Question will result (steps 10. 11, and 13)
AND no Technical Specification revision will be involved.
The change may be implemented in accordance with applicable procedures.
(
) A Technical Specification revision is involved; but no Unreviewed Safety Question will result.
The proposed change requires a License Amendment.
Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.
Mark below as applicable.
(
) The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
(
) The change is a plant modification or minor plant change.
Mark below as applicable.
(
) A revision to an existing Technical Specification is required.
The change HUST NOT be installed until receipt of an approved Technical Specification revision.
(
) The change will not conflict with any existing i
Technical Specifications and only new Technical Specifications are required.
In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization is granted, the block below should be checked.
(
)
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License i
Amendment.
The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only.
l Preparer M. A.
te N/
a Signature Date APPROVED DEC 311990 4100a Q.C.O.S.R,
i 0AP s JC-S21 i
Revision 1 O
15.
The reviewer has determined that the documentation is adequate to l
support the above conclusion and agrees with the conclusion.
t Reviewer
//df-f/
Signature Date 16.
Obtain a safety evaluation number and note at top of page 1.
17.
Forward a copy of this Safety Evaluation to the FSAR Coordinator, Monthly Report Coordinator and Tech Spec Coordinator. (ANI Audit, August 1990).
Completed:
Initial 2/Ar]
Date_ //h2[f/
t O
l s
l i
\\
APPROVED DEC 311990 (final) 4100a Q.C.O.S.R.
i
i 5.
This modification will install an air conditioning system for O
the CRD Repair Room. The air handling unit, plenum, return V
grill and thermostat will be installed inside the CRD repair ante room. The air cooled condensor will be installed outside j
the CRD Repair Room. This equipment will be powered from a 480 V GE MCC which will replace the existing Westinghouse MCC 42R-2-1.
Operation of the CRD Repair Room A/C System will offset the constant addition of heat incurred during maintenance on control rod drives.
The design includes locally mounted disconnect switches for periods when this equipment will not be required.
6.
Possible new failure modes or unacceptable conditions include:
(a)
Electrical failures in the new equipment.
(b)
Leaks in the new refrigerant lines.
(c)
Failures in the modified block walls.
(d)
Spread of contamination.
Possible impact of the above failures during all operating modes are:
(a)
The electrical requircsants for this modification include the installation of properly e.ized breakers in non-safety related MCC 42R-2-1 to protect e?isting plant electrical equipment from any faults which may occur in the new hvac q
equipment. MCC 42R-2-1 receives electrical power from Q
non-safety related transformer T42R-2. The only loads on MCC 42R-2-1 will be the CRD Repair Room HVAC System.
Therefore, a fault in the new electrical equipment will result in the tripping of breakers in MCC 42R-2-1 which will have no impact on any other plant equipment.
(b)
A leak in the refrigerant lines installed by this modification would result in the release of refrigerant-22 into the Unit 1 Reactor Building. The Reactor Building Ventilation System, designed to produce a negative differential pressure, evacuates the Reactor Building at a rate of approximately 1 free volume / hour. Therefore, leakage of refrigerant into the Reactor Building free volume would have no credible impact from a human safety standpoint and have no impact on equipment operation.
(c)
The structural requirements for this modification include design changes to the west (blocking-in an existing louver opening) and north (installation of electrical supply and refrigerant supply and return lines) block walls.
As part of the designer's walkdown, it was identified that no safety related equipment was attached to these two block walls. The actual design will require structural changes meet the seismic 2-over-1 criteria but, if a failure of the wall were to occur, no safety related equipment would be affected.
a
^
(d)
Increased local air flow from the air handling unit could O
result in unacceptable spread of contamination.
The location of the air handling unit inside the ante room instead of the CRD Repair Room provides the-highest air flow in the area of least contamination to prevent an unacceptable airborne contamination problem. Blocking-in the louver opening seals the ante room to prevent the spread of contamination to an uncontrolled area.
l f
.i n
O
)
O D,
QCAP 1000-6 UNIT 1(2)
~
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Chies Nuclear Power Station Date: 5-/ 9-9t/
Reference Nuirden
Subject:
S(R h 9 3 +2d6 Submnted by: /Au CA ed k FOR REVIEWI ~ ' +
Safety Evaluations.tg2Iinvolving an unrevwwed safety question as defined in 10CFR50.59 1.
for:
Changes to procedures as described in the Safety Analysis Report.
a.
J b.
Changes to equipment or systerns as described in the Safety Analysis Report.
l Tests or =mariments.tg2I described in the Safety Analysis Report.
c.
_ 2.
Proposed changes which involve an unroviewed safety ry==Hari as defined in 10CFR50.59.
a.
Procedure changes.
b.
Equipment or system changes.
c,.
Tests or experiments.
3.
Pr=M changes to the Technical Specifications or Operating Ucense.
4.
Nor-4.se wth codes, regulations, orders, Technical SpeclRestions, license requirements, or intamal prwures or instructions having nuclear safety signWicance.
5.
Sivi.;"'we operating abnormalities or deviations from normal and expected p..ieu.-nce of piare aa"i-nent that anects nudear asisty.
6.
M REPORTABLE EVENTS (LERs only).
7.
M i=,ww.M indications of an ur-cr' '
j deficiency in design or Op.u.uvii of safety-related structures, systems, or components.
8.
M changes to the Station Ts v.cy Plan prior to implementation.
9.
M items referred by the Systems Engineering Supervisor. Station Manager, Sne Vice Presidert, and General Manager of Quaky Programs and Assessmerna, frohijiidiillAfi6s:3FET4MEE#e Wh M'"
i j
[
- 10. Other OSR ttoms/ Documents.lH2Iaddressed above.
This Transmittal is being made in accordance wth Quad Cities Nuclear Power Station Technical Si+_i 1-S.1.G.2.d(1) for information only. No specific action is required unless deemed mry by Offste Review and investigative Function.
8 7
Y
e.
Exhibit D ENC-QE-06.1 Revision S Page 1 of 1 1
Technical Specification Revisions for Modification Station Quad Cities Unit (s) 1&2 Modificatien # DCR 4-93-205 4
To:
(Systems Design Superintendent)
J.
Shrace (NLA)
N. Chrissotimos (Station Regulatory Assurance Supervisor) l List required Technical Specification revisions No Technical Specification revision is required as a result of this DCR.
Recommend effective date for re sion (i.e., calendar date, beginning of outage
- , or end of outage #)
i Prepared by: f.. !
. //Vf/
l$'hY
/ UAA "
Dates H
i.-
. /,9h' i
~
1 4
i 3
J
+
h QE-06.1 DECA version 2.3 i
I s
l Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5 Page 1 of 11 i
Station / Unit ouad cities
/ 1&2 i
Exhibit E 10CFR50.59 SAFETY EVALUATION l.
List the documents implementing the proposed change.
DCR 4-93-205, NWR 008212. NWR 008293 2.
Describe the proposed change and the reason for the change.
The torus level indication was found to be in error and the source was traced to the faulty Narrow Range Level Transmitter i
(LT-001-1602-9).
NWR.Q08293 replaced the Rosemount Model 1151DP3B12 (obsolete designation) with a Rosemount Model 1151DP3G12M1Bl. This model includes the mounting bracket which was previously ordered separately and an optional integral meter.
The Reactor Building Exhaust Fan 2C was auto-tripping and the source was traced to the setpoint of differential pressure switch (DPS-002-5741-261C) being at the low er'd of the switch's range.
NWR Q08212 replaced the Dwyer Model 1821-2 with Dwyer Model 1823-1.
This model has a range of 0.3" to 1.0" water column (WC) whereas the previous model's range was 0.5" to 2.0" WC. The setpoint of 0.5" WC remains unchanged.
DCR 4-93-205 updates the appropriate data sheets to reflect these changes.
r 3.
Is the change:
[X) Permanent t
[ ] Temporary -
Expected duration t
AND Plant Mode (s) restrictions while installed I
(NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures,
'i or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation. List any other controlling documents such as SERs, previous modifications or i
Safety Evaluations, etc.
j I
3.4.1.2,
" Internal Flood Measures" 7.5.3,
" Safety Parameter Display System" l
9.4.7,
" Reactor Building Ventilation System" 15.6.5.4.4, " Fission Product Release from Reactor Building to i
Atmosphere" QE-06.1 DECA Version 2.3 I
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Exhibit E Mod # DcR 4-93-205 ENC-QE-06.1 Revision 5 Page 2 of 11 Station / Unit ouad Cities
/j &2 Exhibit E 10CTR50.59 SAFETY EVALUATION 5.
Describe how the change will affect plant operation when the changed SScs function as intended (i.e.,
focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
Torus Narrow Range Level Transmitter:
The "B" designator in the old transmitter model number indicates 10-50 mA DC output.
The "G" designator of new model number indicates 10-50 rA DC output which is the same as the old value and is the same as the value given in the instrument data sheet.
The new transmitter (with bracket and integral meter) weighs 13.34 lb. while the old transmitter and bracket weighs 12 lb and 1.12 lb. = 13.12 lb.
This weight difference is negligible.
The mounting of the transmitter is identical for both transmitter models.
Per the Master Equipment List, Rev. 30, the transmitter is non-EQ.
The Level Transmitter change will not affect plant operation. the new LT has the same accuracy and function as the original. The new LT provides local level indication and a level signal to an indicator and recorder in the control room.
The local display meter is an enhancement.
The LT has no control function.
Reactor Building Exhaust Fan Pressure Switch:
The existing pressure switch is Dwyer Model 1821-2 while the new switch is Dwyer Model 1823-1.
The differences between the new model and the old model are the approving agencies and the instrument range.
1 Model 1821-2 is UL Safety control listed only and Model 1823-1 is UL, CSA and FM approved.
Since there are no commitments for approving agencies, this difference is acceptable.
The old switch, Model 1821-2, has a range of 0.5" to 2.0" QE-06.1 DECA Version 2.3
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' Exhibit E Mod # D'CR 4-93-205 ENC-QE-06.1 l
Revision 5
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Page 3 of 11 Station / Unit Quad Cities
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Exhibit E 10CFR50.59 SAFETY EVALUATION Water Column (WC) with a setpoint of 0.5" WC.
This model will reset at 0.6" WC.
The new switch, Model 1823-1, has a range of 0.3" to 1.0" WC._ The setpoint remains at 0.5" WC and this model will reset at 0.58" WC.
The deadband is slightly narrower for the new switch.
This will have an i
insignificant affect on system operation.
Having the i
setpoint in midrange instead of at the low end will enhance i
system operation by reducing or eliminating spurious. trips.
The Differential Pressure Switch change will not affect' plant operation.
All dimensions, materials and method of operation are the same.
The switch has the same set point j
and function as the original and better accuracy than the original.
6.
Describe how the change will affect equipment failures.
Ir, pnticular, l
describe any new failure modes and their impact during all e X teable operating modes.
[
This change will not affect equipment failures nor will it l
introduce any new failure modes.
The replacement LT has the same l
method of operation, accuracy and performance as the original.
The replacement DPS with a setpoint of 0.5" WC and a range of 0.3" to 1.0" WC will enhance system operation by. reducing or eliminating spurious trips.
7.
, Identify each accident or anticipated transient (i.e., large/small break i
LOCA, loss of load, turbine missiles, fire, flooding) described in the
{
SAR where any of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSc is explicitly or implicitly assumed to functicn during or after the accident Operation or failure of the changed SSC could lead to the accident l
f ACCIDENT SAR SECTION
-Internal Flood 3.4.1.2 Measures-
[
-LOCA 7.5 i
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t Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5
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Page 4 of 11 l
Station / Unit Quad Cities
/ 1&2 i
Exhibit E 10CFR50.59 SAFETY EVALUATION 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be i
affected. To determine the factors affecting the specification, it is-necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.
The torus level measurement requirements are referenced in Sections 3.2/4.2E and 3.7/4.7.A.1 and Tables 3.2-4 and 4.2-2.
l 9.
Will the change involve a Technical Specification revision?
( ) Yes [X) No If a Technical Specification revision is involved, the change cannot lus implemented until the NRC issues a license amendment.
When completing Step 14, indicate that a Technical Specification revision is required.
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Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5 i
Page 5 of 11 Station / Unit ouad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or t
malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7.
Provide the rationale for all No answers.
Af fected accident -Internal Flood
-LOCA SAR Section:
3.4.1.2 7.5 May the probability of the accident be increased?
[ ] Yes [X) No The replacement LT has the same accuracy and function as the original.
The function cef the LT is to provide local level indication and a level signal to an indicator and recorder in the Control Room.
The LT has no control function.
The new DPS provides a faa trip function under no flow conditions.
The old switch has a range of 0.5" to 2.0" Water Column (WC).
Since the setpoint for the switch is 0.5" WC, a i
Model 1823-1 with a range of 0.3 to 1.0 in. WC will enhance i
system operation by reducing or eliminating spurious trips.
Therefore, the pressure switch replacement will not increase the' probability of an accident.
May the consequences of the accident (off-site dose)
[ ] Yes
[X) No be increased?
The new LT bas the same accuracy and function as the original.-
The replacement LT will function as originally designed under all operating and accident conditions.
Thus, the consequences of the accident are not affected by this change.
The new DPS has the same setpoint, accuracy and function as the original.
The replacement DPS will function as originally designed under all operating and accident conditions.
Thus, the consequences of the accident are not affected by this change.
l 1
QE-06.1 DECA Version 2.3 i
Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 f
Revision 5 j
Page 6 of 11 Station / Unit ouad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION T
May the probability of a malfunction of equipment
[ ] Yes
[X) No j
important to safety increase?
The LT functions the same as the original transmitter.
- Thus, this change does not increase the consequences of a malfunction of equipment important to safety.
t The DPS with a setpoint of 0.5" WC and with a range of 0.3 to 1.0" WC will enhance system operation by reducing or eliminating spurious trips.
Therefore, the pressure switch replacement will decrease the probability of a malfunction of equipment important to safety.
May the consequences of a malfunction of equipment
[ ] Yes
[X] No-t important to safety increase?
J The LT and DPS function the same as the original instruments.
Thus, this change does not increase the ccnsequences of a malfunction of equipment important to safety.
If any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
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Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5 Page 7 of 11 Station / Unit Ouad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 11.
Based on your answers to. Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the t
SAR7
[ ] Yes
[X) No Describe the rationale for your answer.
The instrument model changes will not affect the function or 4
operation of the systems since the replacement instruments function the same as the original instruments.
If the answer to Ouestion 11 is Yes, then an Unreviewed Safety Ouestion exists.
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i Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5 Page 8 of 11 Station / Unit Qrad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
3.2/4.2 Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
The desian of the cost-accident instrumentation system and components as described in Section 3.2/4.2 has not been chanced or modified by this DCR.
The carameters used to establish the Technical Specifications have not chanced.
(Check appropriate condition):
[ ] All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction.
Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.
[ ]
The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s)/ margin (s) and applicable reference for the margin of safety below - proceed to question 13.
[ ] The applicable parameter or condition change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit.
Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references.
List the agreed limit (s)/ margin (s) below.
[X) The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to question 14.
List Acceptance Limit (s)/ Margin (s) of Safety Tech Spec SAR Section QE-06.1 DECA Version 2.3 l
Exhibit E l
Mod # DCR 4-93-205 ENC-QE-06.1 l
Revision 5 Page 9 of 11 Station / Unit Quad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION i
SER Section i
4 i
13.
Use the above limits to determine if the margin of safety,is reduced (i.e., the new values exceed the acceptance limits).
Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
1 If a Marcin of Safety is reduced an Unreviewed Safety Question exists.
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. Mod # DCR 4-93-205 ENC-QE-06.1 Revision 5 Page 10 of 11 Station / Unit Quad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 14.
Check one of the following
[ ] An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
[X) No Unreviewed Safety Question will result ( Steps 10, 11, and 13)
AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
[ ] A Technical Specification revision is involved; but no Unreviewed j
Safety Question will result. The proposed change requires a-License Amendment. Notify Station Regulatory Assurance and Nuclear i
Licensing that a Technical Specification revision is required.
Mark below as applicable.
[ ] The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
[ ].The change is a plant modification or minor plant change.
Mark below as applicable.
[ ] A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
[ ] The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required.
In these cases, Nuclear Licensing may 6
authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization is granted, the block below should be checked.
[] Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of' l
the License Amendment. Tne 10CFR50.59 Safety l
Evaluation indicates that no Unreviewed Safety i
Question will result and provides authority-for installation only.
QE-06.1 DECA version 2.3 i
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Exhibit E Mod # DCR 4-93-205 ENC-QE-06.1 l
Revision 5 Page 11 of 11 l
Station / Unit Ouad Cities
/ 1&2 i
Exhibit E 10CFR50.59 SAFETY EVALUATION i
Notes Partial Mo fications and/or separate 10CFR50.59 reviews ~for ti s f the work may be used to facilitate installation.
l Preparer IN/
II M I!/ 9[
i tate
'(Cognidant Engineer) 15.
The reviewer has determined that the documentation is adequate to support the above co lusion and agrees with the conclusion.
Reviewer
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'? N (Design iup[rintendent/ Supe'rvisor)
Date i
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QCAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad C!tles Nudear Power Station Date: 8- /f - H/
Reference Number:
Subject:
5)CR Ll & Q - QLlL n
[/ijer k [t 4 m id [
/
Subm!tted by:
IFOR R5TIEWS ~
Safety EvaluationsE involving an unreviewed safety question as defined in 10CFR50.59 1.
for:
Changes to procedures as described in the Safety Analysis Report.
a.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
Tests or experiments E described in the Safety Analysis Report.
c.
Proposed changes which involve an unreviewed safety question as defined in 10CFM50.59.
2.
s.
Procedure changes.
b.
Equipment or system changes.
c.
Tests or experiments.
3.
Propaavi chanDes to the Technical SE+T-rions or Operating Ucense.
Noncompliance with codes, regulations, orders, Technical Specifications, license 4.
requirements, or intamal prwures or hstrpinns having nuclear safety significance.
Significant operating abnormalities or deviations from normal and expected performance of 5.
plant equipment that affects nuclear ::::8:2y 6.
M REPCRTABLE EVENTS (1.ERs only).
7.
M iswyrdzed indications of an ti u27 --j deficierry in design or operation of safety-related structures, systems, or w,i.,ac rs.
8.
M risriges to the Station Emergency Plan prior to implementation.
M ltems referred by the Systems Engineering Supervisor, Station Manager, Site Vice 9.
President, and General Manager of Qunfity Programs and Assessments.
iFb$505bkl55 I5.?ibh?N$50:A % So ~
[
- 10. Other OSR ltems/DocumentsE addressed abava.
This Tras r.lrsl ls being made in accordance with Quad Cities Nuclear Power Station Technical Spur he 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offs!te Review and inv=Wylve Function.
'8
v Exhibit D ENC-QE-06.1 R0vicion 5 f
Page 1 of 1 Technical Specification Revisions for Modification Station cuad cities Unit (s)
I& 2 Modification # DCR 4-94-046 Standby Gas Treatment (Systems Design Superintendent)
To:
J.
Shrace (NLA)
N. Chrissotimos (Station Regulatory Assurance Supervisor)
List required Technical Specification revisions:
None Recommend effective date for revision (i.e., calendar date, beginning of outage
- , or end of outage )
I Prepared by:
/
. /1 Date:
( /
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DECA Version 2.3 QE-06.1
Exhl. bit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 1 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CTR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
DCR 4-94-046 Standby Gas Treatment (SBGT) System 2.
Describe the proposed change and the reason for the change.
This DCR revises the Master Equipment List (MEL) and selected drawings to incorporate the results of Component Classification (CC) of the Standby Gas Treatment (SBGT) System. As part of this DCR, 1) no physical change was made to any plant structure, system equipment or component and 2) some components were upgraded from NSR to SR because they are required for the SBGT system to perform its SR function (Secondary Containment Radioactive Effluent Control). Documentation specifically addressing these changes is included in Component Classification Binder # CC-QC009. The CC program is an ongoing controlled program that is supervised by Station Engineering.
3.
Is the change:
[X]
Permanent
( ) Temporary -
Expected duration AND Plant Mode (s) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation. List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
6.0.1.4 " Engineered Safeguard Features - Standby Gas Treatment System" 6.2
" Engineered Safeguard Features - Containment Systems" 6.5
" Fission Product Removal and Control Systems" 4
15.6.2
" Break in Reactor Coolant Pressure Boundary Instrument Line Outside Containment" 15.6.5
" Loss of Coolant Accidents Resulting from Piping Breaks Inside Containment" 15.7.2
" Design Basis Fuel Handling Accidents Inside Containment and Spent Fuel Storage Buildings" l
QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 2 of 11 Station / Unit Guad Cities
/1 & 2 Exhibit E 10CTR50.59 SAFETY EVALUATION 5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i.e.,
focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
No physical change was made to any plant system, structure, equipment or component. The component classification program specifically addressed the effect of the drawing and classification changes on the SBGT system safety function and operating modes. Documentation of this is included in the SBGT system component classification binder. Plant operation is not affected by this DCR.
6.
Describe how the change will affect equipment failures.
In particular, describe any new failure modes and their impact during all applicable operating modes.
There were no physical changes made to any plant system, structure, equipment or component by this DCR. The changes documented in this DCR do not create any new operating or failure modes and have no impact on any existing operating or failure modes. Equipment failures are not affected by this DCR.
7.
Identify each accident or anticipated transient (i.e.,
large/small break LoCA, loss of load, turbine missiles, fire, flooding) described in the SAR where any of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or f ailure of tne changed SSC could lead to the accident ACCIDENT SAR SECTION Break in Reactor 15.6.2 Coolant Pressure Boundary Instrument Line Outside Containment Loss of Coolant 15.6.5 Accidents Resultina from Pipina Breaks Inside containment QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 3 of 11 Station / Unit ouad Cities
/1 &2 Exhibit E 10CFR50.59 SAFETY EVALUATION Desian Basis Fuel 15.7.2 Handlina Accidents Inside Containment and Spent Fuel Storace Buildinas 8.
List each Technical Specification (Safety Limit, Limiting Safety System setting or Limiting Condition for operation) where the requirement, associated action items, associated surveillances, or bases may do affected. To determine the factors affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.
The SBGT system and components are described in Technical Specifications Section 3.7/4.7. As part of this DCR, no physical change was made to any plant system, structure, equipment or component. The SBGT component classification process determined that the drawing and classification changes made by this DCR did not alter the safety limits or other parameters used to establish the Technical Specifications. No Technical Specifications are affected by this DCR.
9.
Will the change involve a Technical Specification revision?
[ ] Yes
[X) No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing step 14, indicate that a Technical specification revision is required.
QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 4 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7.
Provide the rationale for all NO answers.
Af fected accident Break in Reactor Coolant Pressure Boundary Instrument Line Outside Containment Loss of Coolant Accident (LOCA)
Resultina from Pinina Breaks Inside Containment Desian Basis Fuel Handlina Accidents Inside Containment and Scent Fuel Storace Buildinas SAR Section:
15.6.2 15.6.5 15.7.2 May the probability of the accident be increased?
{ ] Yes [X) No This DCR does not involve any physical changes to plant systems, structures, equipment, or components. The Component Classification (CC) process evaluated all SBGT system components and identified the operating mode required for each component to accomplish the SBGT system safety function. As a result of the SBGT system CC process, several components were reclassified from NSR to SR. The effect of these classification changes was evaluated through the CC process and was found to have no impact on either the SBGT system safety function or on the accident scenarios analyzed in the UFSAR. The CC process provides assurance that the probability of an accident is not increased i
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QE-06.1 DECA Version 2.3
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Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 5 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION due to the component classification changes. Furthermore, the.CC process provides assurance that these changes do no alter the initial conditions used in any FSAR accident analysis. This CC process is documented in the SBGT system CC binder.
May the consequences of the accident (off-site dose)
[ } Yes
[X) No be increased?
The Component Classification (CC) process evaluated all SBGT system components and identified the operating mode required for each component to accomplish the SBGT system safety function and to mitigate the accidents analyzed in the UFSAR. As part of the CC process, several components were reclassified from NSR to SR.
These classification changes were evaluated through the CC process and were found to have no impact on the SBGT system's l
ability to mitigate the affects of an accident. The CC process provides assurance that the consequences of an accident are not increased due to the changes in component classification. This CC process is documented in the SBGT system CC binder.
Hay the probability of a malfunction of equipment
[ ] Yes (X) No important to safety increase?
The SBGT system Component Classification (CC) process considered.
all possible equipment and component malfunctions in determining the classification of each SBGTS system component. As part of the CC process, several components were reclassified from NSR to SR.
l The classification changes were evaluated through the SBGT system l
CC process and were found not to have any impact on the SBGT system. The CC process provides assurance that the probability of a malfunction in equipment important to safety is not increased as a consequence of the component classification changes. The-function of the SBGT system and its ability to operate are unchanged.
May the consequences of a malfunction of equipment
[ } Yes
[X) No
.important to safety increase?
The Component Classification (CC) process identified the operating and failure modes of all SBGT system components and their role in accomplishing the SBGT system safety function. As part of CC process, several components were reclassified from NSR to SR. These classification changes were evaluated through the CC process and were found to have no impact on the SBGT system. The CC process provides assurance that the consequences of a malfunction in equipment important to safety are not increased QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 6 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION due to the changes made by this DCR. Results of the CC process for the SBGT system are recorded in the CC binder for the SBGT system.
If any answer to Question 10 is YES, then an Unreviewed safety Question exists.
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Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Cas Treatment Revision 5 Page 7 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 11.
Based on your answere to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a typo different from those evaluated in the SAR7
[ ] Yes
[X] No Describe the rationalo for your answer.
This DCR does not involve any physical changes to plant systems, structures, equipment or components. The Component Classification (CC) process for the SBGT system identified the operating mode for each component in the system and also identified that component's role in accomplishing the SBGT system safety function. The CC process also considered all applicable accidents analyzed in the SAR and all potential equipment or component malfunctions. The CC process provides assurance that the changes made by this DCR do not affect any existing accidents analyzed in the SAR and do not create any new accidents. The SBGT system CC process is documented in the SBGT system CC binder.
If the answer to ouestion 11 is Yes, then an Unreviewed Safety Ouestion exists.
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QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Revision 5 Standby Gas Treatment Page 8 of 11 Station / Unit Quad Cities
/1 & 2 Exhibit E 10CTR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
Technical Specification Section 3.7/4.7 Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
The desian basis for the SBGT System and components. as described in Technical Specification section 3.7/4.7. has not been chanced or modified by this DCR. The parameters and limits used in the Technical Specifications are not chanced.
(Check appropriate condition):
[ ] All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction.
Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.
[ ]
The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition.
List the limit (s)/ margin (s) and epplicable reference for the margin of safety below - proceed to question 13.
[ ] The applicable parameter or condition change is in a pote.tially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit.
Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references.
List the agreed limit (s)/ margin (s) below.
[X) The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety.
Proceed to question 14.
List Acceptance Limit (s)/ Margin (s) of Safety Tech Spec SAR Section QE-06.1 DECA Version 2.3 l
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Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Cas Treatment Revision 5 Page 9 of 11 Station / Unit Quad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION SER Section 13.
Use the above limits to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
1 If a Marain of Safety is reduced an Unreviewed Safety Question exists.
i IJ A l
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QE-06.1 DECA Version 2.3
4 Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 10 of 11 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 14.
Check one of the following:
( ) An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
Tha proposed change MUST NOT be implemented without NRC approval.
[X] No Unreviewed Safety Question will result ( Steps 10, 11, and 13)
AND no Technical Specification revision will be involved.
The change may be implemented in accordance with applicable procedures.
[ ] A Technical Specification revision is involved; but no Unreviewed Safety Question will result.
The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.
Mark below as applicable.
[ ] The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59.
Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
[ ] The change is a plant modification or minor plant change.
Mark below as applicable.
[ ] A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
[ ] The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization is granted, the block below should be checked.
[ ] NucAear Licensing has authorized installation, but j
not operation, prior to receipt of NRC approval of j
the License Amendment.
The 10CFR50.59 Safety i
Evaluation indicates that r.a Unreviewed Safety Question will result and provides authority for installation only.
QE-06.1 DECA Version 2.3 i
Exhibit E Mod # DCR 4-94-046 ENC-QE-06.1 Standby Gas Treatment Revision 5 Page 11 of 11 Station / Unit Ouad cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION artiap^fodificationsand/orseparate10CFR50.59reviewsfor Notes o tie s of the work may be used to facilitate installation.
Preparer # a 5/5/9'/
(Cogbizant E'ngineer)
/ /Date 15.
The reviewer has determined that the documentation is adequate to support the above conclusion and agrees with the conclusion.
Reviewer I ff
(' Design superintendent / Supervisor)
Date i
l QE-06.1 DECA Version 2.3 I
o m 1000-6 i
UNIT 1(2) arvIsION o ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Chies Nuclear Power Staten Date:
5-6-94 Reference Nima,s.; -
Sc2M-b0R S W-055 n
n Submitted by: [ Milt c, Y _ b e di I
s
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TOR REVIEW:.,
- ~
Safety EvaluationsE invahnng an unreviewed safety question as defined in 10CFR50.59 1.
W Changes to procedures as described in the Safety Analyss Report.
a.
b.
Changes to equipment or systems as desenbod in the Safety Analysis Report.
i Tests or==Iments.tHZ[ described in the Safety Anstysis Report.
c.
I 4
Proposed changes which involve an unroviewed asisty question as defined in 10CFR50.59.
2.
i a.
Procedure changes.
l b.
Equipmert or system changes.
c.
Tests or experiments.
3.
Pie-:=i changes to the Technical Specmastions or Operating Uconse.
1 4.
Noncompliance wkh codes, regtdations, orders Technical SpecEcstions, license requirements, or intamal procedures or instructions having nuclear safety significance.
l Significart operating abnormanties or deviations from normal and expected performance of 5.
plant sculpment that anects nudear assoty.
6.
All REPORTABLE EVENTS (LERs only).
7.
AM recogntred indications of an unarP=f 4.~.
iin design or operation of safety-f reisted structures, systems, or components.
8.
AB changes to the Station E.,
i.i lan prior to implementation.
P I
9.
AB ltoms referred by the Systems Engineering Supervisor, Sistion Manager, See Vlos President, and Genomi Manager of OumDty Programs and Assessmerna.
fi6h"5$5lElbificiliiGEEP5MkisarkNi%e:0
- 10. Other OSR ttoms/DocumentsE addressed above.
3 This Transmittal is being made in accordance with Quad Cities Nudear Power Station l
Technkal RM 6.1.G.2.d(1) for information only. No specific action is required unless doomed necessary by Offste Review and leW Function.
i 1
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Exhibit D 1
ENC-QE-06.1 Revision 5 Page 1 of 1 Technical specification Revisions for Modification Station Ouad Cities Unit (s) 1&2 Modification # DCR 4-94-055 (Systems Design Superintendent)
Tor i
J.
Shrace (NLA)
N. Chrissotimos (Station Regulatory Assurance Supervisor)
List required Technical Specification revisions:
There are no required revisions to the Technical Specifications as a result of this change.
Recommend effective date for revision (i.e., calendar date, beginning of outage
- , or end of outage #
8 NI
,t Dates Prepared by I
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Exhibit E Mod # DCR 4-94-055 ENC-QE-06.1 Revision 5 Page 1 of 12 Station / Unit ouad cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
DCR 4-94-055 2.
Describe the proposed change and the reason for the change.
The implemented change will incorporate the actual location of pressure test point connections for the condensate booster pump discharge piping on Unit 1 Piping and Instrumentation Diagram (P&ID); and incorporate the addition of pressure test point connection for the condensate booster pump discharge piping on Unit 2 P&ID. These Unit 1 and Unit 2 P&ID as-built changes reflect the original designed and installed conditions.
3.
In the change:
[X)
Permanent
( )
Temporary -
Expected duration AND Plant Mode's) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation. List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
Functional aspects or activities related with the Condensate or Condensate Pump Room are described in the following UFSAR Sections:
1.2.2.2,
" Station Arrangements" 3.4.1.2.1,
" Protection of the Condensate Pump Room and Residual Heat Removal Service Water Pump Rooms" 3.6.1.1.2, "High Energy Systems" 5.1.3,
" Reactor Coolant System Subsystems" 5.4.7.2.3, "Other Functions of the Residual Heat Removal System" 6.3.3.2.6,
" Summary - Integrated Emergency Core Cooling System Performance Evaluation" i
6.3.3.2.8.1, "Small Line Break" 7.7.6, " Main Condenser, Condensate and Condensate Demineralizer" 9.2.8.2,
" System Description - Standby Coolant Supply System" QE-06.1 DECA Version 2.3
Exhibit E ENC-QE-06.1 Mod # DCR 4-94-055 Revision 5 Page 2 of 12 Station / Unit ouad Cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 10.1, " Summary Description - Steam and Power Conversion System" 10.4.7, " Condensate and Feedwater System" Table 10.4-3, " Condensate Booster Pump Characteristics" 11.1.3.7,
" Tritium" Table 11.1-7, " Turbine Building Equipment Drain Sump Sources For Radioactive Material" Table 12.3-3, " Quad Cities Unit 1 Area Radiation Monitoring System Sensor Location and Range" Table 12.3-4, " Quad Cities Unit 2 Area Radiation Monitoring System Sensor Location and Range" 14.2.12.1.32.2, " Condensate and Feedwater Systems" i
UFSAR Accident Analysis Sections Pertaining to Condensate:
15.8, " Anticipated Transients Without SCRAM" Other Documents:
Grinnell Erection drawing number 1-1401-ED-1 Revision 01/30/70 Grinnell Erection drawing number 2-3401-ED-1 Revision 08/31/71 Quad Cities Special Report 3A - Condensate Pump Room Modifications (Permananet Flood Protection of the RHR Service Water Pumps and Diesel Generator Cooling Water Pumps) 5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
The relocation or addition of the Unit 1 and Unit 2 condensate booster pump discharge pressure test point connection does not produce any functional change in the system.
It only revises the Unit 1 and Unit 2 P& ids to reflect the original designed and installed conditions for pressure testing tap points.
Implementation of these changes will not alter any operational parameters of the system or the plant, and therefore will not affect current plant operation.
QE-06.1 DECA Version 2.3
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t Exhibit E ENC-QE-06.1-Mod # DCR 4-94-055 Revision 5 Page 3 of 12 Station / Unit ouad Cities
/ IG2 3
Exhibit E 10CFR50.59 SAFETY EVALUATIoK i
6.
Describe how the change will affect equipment failures.
In particular, l
describe any new failure modes and their impact during all applicable j
operating modes.
t This change does not add any new components to the system, but l
reflects actual pressure tap locations which were installed in 1970 and 1971 for Units 1 and 2, respectively, as shown on the original Grinnell erection drawings.
The operational characteristics of the system will not be affected by this P&ID drafting change, so there is no potential for introduction of any
~
circumstances or conditions that could produce a failure mechanism that did not previously exist.
7.
Identify each accident or anticipated transient (i.e., large/small break LoCA, loss of load, turbine missiles, fire, flooding) described in the SAR where any of the following is truet The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function t
during or after the accident Operation or failure of the changed SSc could lead to the accident ACCIDENT SAR SECTION t
7
-Loss of Normal AC 15.8.2 Power
~-f_,oss of Normal 15.8.3 Feedwater Flow 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine the factors affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.
The applicable Safety Limits, Limiting Safety System Settings and Limiting Conditions for Operation are not directly related to, nor do they mention the Condensate System piping and valves.
Therefore, no Technical Specifications require revision as a result of this change.
The effects from possible failure of l
condensate piping are described in Technical Specifications Sections 3.5/4.5 and 3.9/4.9.
However the limiting conditions l
I stated for condensate pump room flood protection and liquid radioactive effluents are not affected.
T QE-06.1 DECA Version 2.3 l
4 Exhibit E ENC-QE-06.1 Mod # DCR 4-94-055 Revision 5 Page 4 of 12
/_1&2 Station / Unit Ouad Cities Exhibit E 10CFR50.59 SAFETY EVALUATION Will the change involve a Technical Specification revision?
9.
[ ] Yes [X) No If a Technical specification revision is involved, the change cannot be When completing implemented until the NRC issues a license amendment.
indicate that a Technical specification revision is required.
step 14, QE-06.1 DECA version 2.3
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Exhibit E ENC-QE-06.1 I
Mod # DCR 4-94-055 Revision 5' Page 5 of 12.
/ 1&2 Station / Unit ouad Cities
?
Exhibit E 10CFR50.59 SAFETY EVALUATION To determine if the probability or the consequences of an accident or 10.
ma* function of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7.
Provide the rationale for all No answers.
Affected accident Loss of Normal AC Power i
SAR Section:
15.8.2 May the probability of the accident be increased?
[ } Yes [X] No
?
i i
The probability of a Loss of Normal AC Power event is independent of the function or operation of the Condensate System. This i
change can not increase the probability of the initiating event for the Loss of Normal AC Power.
i May the consequences of the accident (off-site dose)
{ ] Yes [X) No be increased?
i The loss of Normal AC Power would deenergize the condensate Therefore, the possiblity of this change affecting system pumps.
The condensate pump operation has previously been analyzed.
potential consequences of this accident which affects system operation and off-site dose are not increased.
May the probability of a malfunction of equipment
[ ] Yes
[X] No important to safety increase?
)
The incorporation of the originally designed and installed pressure test connections will not increase the probability of a malfunction of equipment important to safety due to Loss of The affects of Loss of Normal AC Power has Normal AC Power.
previously been analyzed which would deenergize the normally operating equipment including the condensate pump system.
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Exhibit E i
ENC-QE-06.1
. Mod # DCR 4-94-055 Revision 5 Page 6 of 12 Station / Unit Ouad Cities
/ 1&2 i
Exhibit E 10CN50.59 SAFETY EVALUATION i
May the consequences of a malfunction of equipment
[ ] Yes [X} No important to safety increass?
The change as described will not_ affect any operational parameters of the Condensate system.
The consequences of a malfunction of equipment important to safety affected by the Loss of Normal AC Power will not increase.
If any answer to Question 10 is YES, then an Unreviewed Safety Ouestion exists.
I i
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Exhibit E~
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Mod # DCR 4-94-055 Revision 5 l
Page 7 of 12
/ 1&2 Station / Unit Quad cities Exhibit E 10CFR50.59 SAFETY EVALUATION
^
Af fected accident Loss of Normal Feedwater Flow SAR Section:
15.8.3 v
May the probability of the accident be increased?
[ } Yes
[X) No l
The probability of a Loss of Normal Feedwater Flow will not increased due to the incorporation of the originally designed and l
installed pressure test connections.
l May the consequences of the accident (off-site dose)
( ) Yes [X) No be increaced?
The loss of condensate resulting from system failure would result in Loss of Normal Feedwater Flow.
This loss of feedwater flow I
has previously been analyzed.
The potential consequences of this accident which affects system operation and off-site dose are not l
1 increased.
?
May the probability of a malfunction of equipment
[ ] Yes
[X) No important to safety increase?
This change will not increase the probability of a malfunction of equipment important to safety due to Loss of Normal Feedwater Flow.
Failure of this change could result in condensate pump room flooding, however this change does not involve any new components therefore the probability is not increased.
l C
May the consequences of a malfunction of equipment
[ ] Yes
[X) No important to safety increase?
The change as described will not affect any operational j
parameters of the Condensate System.
The consequences of a t
malfunction of equipment important to safety'affected by the Loss of Normal Feedwater Flow or condensate pump room flood will.not i
increase.
The effects and preventative measures of a condensate pump room flood has previously been analyzed for its effects on
[
l equipment important to safety.
If any answer to ouestion 10 is YES, then an Unreviewed Safety Ouestion exists.
i t
QE-06.1 DECA version 2.3 i
- - ~ -. -
T Exhibit E j
Mod # DCR 4-94-055 ENC-QE-06.1 Revision 5 Page 8 of 12 i
Station / Unit ouad Cities
/ 1&2 l
Exhibit E 10CFR50.59 SAFETY EVALUATION 1
i 11.
Based on your answers to Questions 5 and 6, does the change adversely I
impact systems or functions so as to create the possibility of an i
accident or malfunction of a type different from those evaluated in the SAR7 t
( ) Yes (X) No Describe the rationale for your answer.
{
i The change as described does not cause a functional change in the
[
system or its interaction with other plant systems.
It does not alter any physical parameters or process variables of the plant.
Due to the nature of the change, there are no new inherent failure modes introduced to the system and the change does not f
add any new components or process routes.
i If the answer to ouestion 11 is Yes, then an Unreviewed Safety Question i
exists.
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Exhibit E Mod # DCR 4-94-055 ENC-QE-06.1 Revision 5 Page 9 of 12 Str. tion / Unit ouad Cities
/ IG2 Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
Technical Specification Sections 3.5/4.5 and 3.9/4.9.
Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
The effects from oossible failure of the Condensate System and components as described in Technical Soecification Section
^
3.5/4.5 and 3.9/4.9 have not been chanced or modified by this DCR. The carameters and limits used in the Technical Specifications are not chanced.
(Check appropriate condition):
( } All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need~not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.
[ ] The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition.
List the limit (s)/ margin (s) and applicable reference for the margin of safety below - proceed to question 13.
[ ] The applicable parameter or condition change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit. Request Nuclear Licensing assistance to identify the e
acceptance limit / margin for the Margin of Safety determination by
[
consulting the NRC, SAR, SER's or other appropriate references.
l List the agreed limit (s)/ margin (s) below.
l
[X) The change does not affect any parameters upon which Technical I
Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to questir1 14.
i List Acceptance Limit (s)/ Margin (s) of Safety 1
Tech Spec l
QE-06.1 DECA Version 2.3
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c Exhibit E Mod # DCR 4-94-055 ENC-QE-06.1 l
Revision 5 Page 10 of 12 Station / Unit ound cities
/ 1&2 i
i i
Exhibit E t
10CFR50.59 SAFETY EVALUATION SAR Section SER Section r
13.
Use the above limits to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a deceription of compensating factors used to reach that conclusion.
If a Marcin of Safety is reduced an Unreviewed Safety Question exists.
i 1
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]
ENC-QE-06.1 Mod # DCR 4-94-055 Revision 5 l
Page 11 of 12 Station / Unit Quad cities
/ 1&2 Exhibit E 10CFR50.59 SAFETY EVALUATION 14.
Check one of the following I
[ ] An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
(
[X) No Unreviewed Safety Question will result ( Steps 10, 11, and 13)
AND no Technical Specification revision will be involved. The j
change may be implemented in accordance with applicable procedures.
l
[ ] A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a t
License Amendment.
Notify Station Regulatory Assurance and Nuclear l
Licensing that a Technical Specification revision is required.
Mark below as applicable,
[ ] The change is not a p.... modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the the approved Technical Specification change from the NRC, change may be implemented.
[ ] The change is a plant modification or minor plant change.
Mark below as applicable,
[ ] A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
[ ] The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required.
In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization is granted, the block below should be checked.
[ ] Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only.
1 i
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Exhibit E
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ENC-QE-06.1 Mod # DCR 4-94-055 Revision 5 Page 12 of 12
/ 1&2 Station / Unit ouad Cities
)
Exhibit E 10CFR50.59 SAFETY EVALUATION Partial J4odifications and/or separate 10CFR50.59 reviews for Notes f the work may be used to facilitate installation.
t on Y M7Y Preparer
- / Date (CognN, ant Engineer) l
{
The reviewer has determined that the documentation is adequate to 15.
support the above con lusion and agrees with the conclusion.
/!fV J
r Reviewer Date
( De s ign 'Supefintende nt / Supe rvi so r )
C e
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a DECA Version 2.3 QE-06.1 l
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i gcAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station Date:
/$-b C///
Reference Number:
See BCR 4/- W-0 64 n
Submitted by: [ NIrc (M 4 vaedi soR RwiEws;;r Safety EvaluationsE involving an unreviewed safety question as defined in 10CFRSO.59 1.
W Changes to procedures as described in the Safety Analysis Report.
a.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
Tests or =martmentsM described in the Safety Analysis Report.
c.
_ 2.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59.
a.
Procedure changes.
b.
Equipment or system changes.
c.
Tests or suportnants.
3.
PW changes to the Technical Specmastions or Operating Ucones.
4.
Noncompliance wth codes, regtdations, oniers, Technical SpecElcations, license requirements, or intamal procedures or instructions having nuciesr asfety.;v..;"
r.cs.
5.
SignEcant operating abnormalities or deviations from normal and expected performance of plant egulpment that allects nudear asisty.
a.
M REPORTABI.E EVENTS 6.ERs only).
7.
M recogntand indications at an unse d.r :
y in design or +.;;. of safety-reisted structures, systems, or components.
8.
M changes to the Station I.T-v y Plan prior to implementation.
9.
M kams relemed by the Systems Engineering Supervisor, Station Manager, Site Vice President, and sensral Manager or ausmy Programs and Assessments.
INDEiE55ft55fdfE5E$hkb5dNSEMNidNO
[ 10. Other OSR hems /DocumentsE addressed above.
This Transmutal is being rnede in accordanos with Quad Cities Nuclear Power Station Technical AM 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and irwestigative Function.
s
~
Exhibit D ENC-QE-06.1 Revision 5 i
Page 1 of 1 Technical Specification Revisions for Modification Station ouad Cities v
Unit (s) 1&2 Modification # DCR 4-94-064
[
To:
(Systems resign Superintendent)
J.
Shrace (NLA)
N. Chris ec t irnos (Station Regulatory Assurance Supervisor)
Ii List required Technical Specification revisions:
A revision to the Technical Specifications is not required.
Recommend effective date for revision (i.e.,
calendar date, beginning of outage
- , or end ! outage #)
8[ /f Prepared by:' # M
/
dAL Date
/
ll
/ /
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E i
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t 1
QE-06.1 DECA version 2.3
Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 1 of 10 Station / Unit ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
DCR 4-94-064 2.
Describe the proposed change and the reason for the change.
This DCR was submitted to document the following as-builts:
Schematic Diagrams 4E-1351B, Sheet 2; 4E-2345, Sheet 1; 4E-2345, Sheet 2; 4E-2430, Sheet 2; and 4E-2430, Sheet 4:
These drawings update cross references and descriptions on relays and control contacts to more accurately reflect the installed conditions.
Piping Diagram M-84, Sheet 1:
This drawing revises the Equipment Piece Number (EPN) for the Unit 2A Off-Gas Filter Outlet Valve from 2-5499-55 to 2-5499-51.
This change is being made to match the configuration and numbering of the Unit i valve.
3.
Is the change:
[X]
Permanent
[ ] Temporary -
Expected duration AND Plant Mode (s) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation.
List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
5.2,
" Integrity of Reactor Coolant System" 6.0.1.5,
" Emergency Core Cooling System" 6.2,
" Containment Systems" 6.3,
" Emergency Core Cooling Systems" 7.3,
" Engineered Safety Features" 8.0,
" Electric Power" 8.2, "Offsite Power Systems" 8.3, "Onsite Power Systems" 9.5, "Other Auxiliary Systems" 10.4.2,
" Main Condenser Evacuation System" 11.3,
" Gaseous Waste Management System" j
QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 2 of 10 Station / Unit ouad Cities
/1 &2 Exhibit E 10CFR50.59 SAFETY EVALUATION 11.5,
" Process & Effluent Radiological Monitoring & Sampling Systems" 15.2.2.1 " Load Rejection (Generator Trip) Without Bypass" 15.2.2.2 " Load Rejection With Bypass" 15.6
" Decrease in Reactor Coolant Inventory" 5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i.e.,
focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
Updating the cross references on the Schematic Diagrams and the EPN on the Piping Diagrams to reflect the actual plant conditions will not impact any plant system.
Updating the drawings will simplify operations and maintenance on the systems.
6.
Describe how the change will affect equipment failures.
In particular, describe any new failure modes and their impact during all applicable operating modes.
Updating the Schematic Diagrams to correct cross references and relay designations on the Diesel Generator and Core Spray systems and revising the EPN on the Unit 2A Off-Gas Filter Inlet Valve to match the Unit 1 configuration does not introduce any new failure modes in these systems.
7.
Identify each accident or anticipated transient (i.e.,
large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the SAR where any of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function during or after the accident operation or f ailure of the changed SSC could lead to the accident ACCIDENT SAR SECTION
-Loss of auxiliary 8.3.1 power
-Power bus loss of 8.3.1 voltace
-Failure of one 8.3.1.6.4 diesel aenerator to start
-Load reiection 15.2.2.1 without bvoass QE-06.1 DECA Version 2.3
Exhibit E Mod # DcR 4-94-064 ENC-QE-06.1 Revision 5 Page 3 of 10 Station / Unit Ouad cities
/J & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION
-Load reiection 15.2.2.2 with bypass (Loss of electrical load)
-Loss of Coolant 15.6.2.
15.6.5 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting condition for operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine the factors affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.
The Core Spray system is addressed in the Technical Specifications (Tech Spec) Section 3.5/4.5. The Diesel Generator system is referenced in Tech Spec Section 3.9/4.9.
Updating the drawing cross references and descriptions on relays and control contacts does not impact the Tech Specs.
The Off-Gas system is referenced in Tech Spec Section 3.8/4.8.
Revising the EPN number on the drawing for the Unit 2A Off-Gas Filter Outlet Valve does not impact the Tech specs.
9.
Will the change involve a Technical Specification revision?
( } Yes
[X} No If a Technical specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing step 14, indicate that a Technical specification revision is required.
i i
l QE-06.1 DECA Version 2.3
Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 4 of 10 Station / Unit ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or malfunction of equipment imprtant to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7.
Provide the rationale for all No answers.
Affected accident -Loss of auxiliary cover
-Power bus loss of voltaae
-Failure of one diesel cenerator to start
-Load reiection without bvoass
-Load reiection with bvoass (Loss of electrical load)
-Loss of coolant SAR Section:
8.3.1 8.3.1 8.3.1.6.4 15.2.2.1 15.2.2.2 15.6.2. 15.6.5 May the probability of the accident be increased?
[ ] Yes
[X) No The function of the Core Spray and the Diesel Generator Systems are unchanged by updating the drawing cross references and descriptions on relays and control contacts.
Likewise, the function of the Off-Gas System and its ability to operate are unchanged by the revision to the EPN on the Unit 2A QE-06.1 DECA Version 2.3
+
Exhibit E Mod / DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 5 of 10 Station / Unit Ouad Cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION Off-Gas Filter Outlet Valve. Information on this label required for updating station procedures is being coordinated by the station system engineers.
The revised EPN on the drawing will provide consistency with the Unit 1 Off-Gas system.
May the consequences of the accident (off-site dose)
[ ] Yes (X) No be increased?
The function of the Core Spray, Diesel Generator and Off-Gas Systems and their ability to operate are unchanged by the revision to the cross references and descriptions on relays and control contacts on the Schematic Diagrams and the change in EPN for the Unit 2A Off-Gas Filter Outlet Valve (2-5499-51). There is no change in any accident scenarios and no new failure modes are introduced by these changes.
May the probability of a malfunction of equipment
( ) Yes (X) No important to safety increase?
The probability of equipment malfunction is unchanged because there is no physical change to the equipment or operating modes by revising the cross references and descriptions on relays and control contacts on the schematic diagrams or revising the EPN on the P&ID.
Operations and maintenance will be enhanced by these revisions.
May the consequences of a malfunction of equipment
[ ] Yes
[X) No important to safety increase?
The probability of malfunction of any equipment or system due to the updating of the cross references and descriptions on relays and control contacts and by revising the EPN for valve 2-5499-51 is not increased and therefore the consequences of a malfunction of equipment important to safety are not increased. All systems will function as originally designed.
If any answer to Ouestion 10 is YES, then an Unreviewed Safety Ouestion exists.
QE-06.1 DECA Version 2.3
. ~.
l Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1-Revision 5 l
Page 6 of 10 i
Station / Unit Quad cities
/1 & 2 I
Exhibit E l
10CFR50.59 SAFETY EVALUATION 11.
Based on your answers to Questions 5 and 6, does the change adversely i
impact systems or functions so as to create the possibility of an i
accident or malfunction of a type different from those evaluated in the SAR7
[ ] Yes
[X) No l'
Describe the rationale for your answer.
No new accident scenarios are created by this DCR.
The function of the Core Spray, Diesel Generator and Off-Gas Systems and their ability to operate are unchanged.
This DCR will not adversely impact systems or functions nor will the possibility of an accident malfunction be created that is different from those i
previously evaluated in the SAR.
If the answer to ouestion 11 is Yes, then an Unreviewed safety ouestf.g 2
exists.
I i
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QE-06.1 DECA Version 2.3 i
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Exhibit-E
' Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 7 of 10 Station / Unit Quad cities
/1 & 2 Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
Technical Specification Sections 3.5/4.5. 3.8/4.8 and 3.9/4.9.
Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
The carameters used to establish the Technical Soecifications for the Core Sorav. Diesel Generator and Off-Gas systems are not chanced by this DCR. This DCR updates cross references and descriotions on relavs and control contacts on the Core Sorav and Diesel Generator systems and relabels eculoment associated with the Off-Gas Filter Outlet Valve.
(Check appropriate condition):
[ ] All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.
[ ]
The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition.
List the limit (s)/ margin (s) and applicable reference for the margin of safety below - proceed to question 13.
[ ]
The applicable parameter or condition change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit. Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references.
List the agreed limit (s)/ margin (s) below.
[X) The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the i
margin of safety. Proceed to question 14.
List Acceptance Limit (s)/ Margin (s) of Safety Tech Spec QE-06.1 DECA Version 2.3
=..
i' Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 8 of 10 Station / Unit Quad Cities
/1 & 7 f
Exhibit E i
10CFR50.59 SAFETY EVALUATION SAR Section SER Section 13.
Use the above limits to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
If a Marain of safety is reduced an Unreviewed safety Question exists.
t3/A E
QE-06.1 DECA Version 2.3 i
t I
Exhibit E l
Mod #'DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 9 of 10 Station / Unit Quad cities
/1 & 2 l
[
?
Exhibit E l
10CFR50.59 SAFETY EVALUATION 14.
Check one of the followings i
[ ] An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
[X)
No Unreviewed Safety Question will result ( Steps 10, 11, and 13) i AND no Technical Specification revision will be involved. The 1
change may be implemented in accordance with applicable procedures.
j
[ ]
A Technical Specification revision is involved; but no Unreviewed
[
Safety Question will result. The proposed change requires a
[
License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.
~
Mark below as applicable.
[ ] The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59.
Upon receipt of the approved Technical Specification change from the NRC, the i
change may be implemented.
l
[ ] The change is a plant modification or minor plant change.
Mark below as applicable.
[
i
[ ] A revision to an existing Technical Specification is l
required. The change MUST NOT be installed until receipt i
of an approved Technical Specification revision.
{
[ ] The change will not conflict with any existing Technical l
Specifications and only new Technical. Specifications are required.
In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If
{
such authorization is granted, the block below should be checked.
i
[ ] Nuclear Licensing has authorized installation, but j
not operation, prior to receipt of NRC approval of the License Amendment.
The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety l
Question will result and provides authority for j
installation only.
j t
f i
QE-06.1 DECA Version 2.3 i
- a *. ',i Exhibit E Mod # DCR 4-94-064 ENC-QE-06.1 Revision 5 Page 10 of 10 l
station / Unit ouad cities
/1 & 2
)
)
Exhibit E 10CFR50.59 SAFETY EVALUATION Notes Partial Modifications and/or separate 10CFR50.59' reviews for port ns the work may be used to facilitate installation.
Preparer M23
/l b 8N/II (6cgniza/t E'ngineer)
Date 15.
The reviewer has determined that the documentation is adequate to support the above conc usion and agrees with the conclusion.
[
M Reviewer (Design'up'er[ntendent/ Supervisor)
Date S
i
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GCAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 Of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMI
(.-
t Quad CPJes Nuclear Power Station Date:
f- / 7-fly' Reference Ntmd.si; sST-73
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mW Safety Evaluations,t(QIinvoMng an unreviewed safety question as denned in 10CFR50.59 1.
for:
Changes to procedures as described in the Safety Analysis Report.
a.
Changes to equipment or systems as described in the Safety Analysis Report.
b.
Tests or W.tg described in the Saisty Analysis Report.
c.
Proposed changes which irwolve an untsviewed asisty question as defined in 10CFR50.50.
2.
s.
Procedure changes.
b.
Equipment,or system changes.
f, c.
Tests or exportments.
t 3.
Proposed changes to the Technical AN or Operating License.
Noncompliance with codes, regidstions, orders, Technical SpeclRcations, license 4.
requirements, or intamal prM=es or instructions having nutdaar asisty significance Signincert operating abnormouties or deviations from normal and awpar*ad performance of 5.
piart equipment that a5eces nutiesr asisty.
a.
M REPORTABl.E EVENTS (LERs only).
M recognized indications of an unanticipstod danciency in design or operation of safety-7.
reisted structures. synsms. or componsraa. -
a.
M chenpas to the Station Emergency Plan prior to imalamentation.
M koms refened by the Systems Engineering Supervisor, Station Manager, Ste Vice 9.
President, and General Manspor of QuaRy Programs and Asassamorts.
!PORINFORMATION:__---- MMI6df*SAWu....,..-... ds""W w n,-.
eMW
)
- 10. Other OSR hems / Documents.tg addressed abtwo.
This Tranammel is being made in accordance wth Ound Cities Nucdear Power Station Technical Aparer=*irvis a.1.G.2.d(1) inr information only. No speclRc action is required unless deemed necessary by Offslie Review anti irwestigadve Function.
8
.e 5
a
3s CGE QAP 1100-S21 Revision 3 10CFR50.59 SAFETY EVALUATIONS January 1993 Safety Evaluation Number: SE-93
-109 Document Identifier: M04-1(21-89-115 work nackaces 002824. 002825 Elect. Mech. Inst (Modification, Temp Alt, Work Request Number, etc.)
Unit (s): 1(2)
System (s):
1700.3900.4200 l
Applicable Plant Mode (s): This evaluation is anolicable for all clant modes.
(RUN, STARTUP/ HOT STNBY, REFUEL or SHUTDOWN)
Plant Mode Restriction (s): This evaluation contains no mode restrictions.
1.
Describe the proposed change:
The work to be performed under this package will calibrate two 0-100 psi pressure indicators (Sl# 208039) and a flow switch (SI# 699224) prior to installation of modification M04-1(2)-
89-115 (Modification of the service water radiation monitoring system sample delivery piping). The pressure indicators (PIs) will be used to ensure the sample stream eductor is operating properly. Under this package, the PIs will be used to gather system sample pressures while throttling the two globe valves on either side of the eductor. Also, during this test, a flow indicator will be installed to give flow indications. This information will be used to determine proper system operating pressures.Jhe.Jiow indicator will then be removed and the flow switch will be installed. The low flow setpoint will then be verified. If erratic indication occurs during performance of the traveler, individual instrument calibrations can be performed.
2.
Reason for the change:
This work package was written to install the instrument portion of modification M04-1(2) 115.
3.
Is the change:
l l
( X ) Permanent j
(
) Temporary - Expected Duration:
]
o:\\CGEiQAP\\l100\\l100-521 CGE QAP 1100-S21 Revision 3 4
List the reference documents reviewed which describes the structure, system or component.
(Identify documents referenced even if no information was found in that section.)
a.
UFSAR Section(s): Table 1.8-1. 3.0. 7.1. 7.5. 11.5.2.7.15 b.
SER Section(s): -None c.
Tech Spec Section(s): 3.2/4.2. 3.8/4.8 d.
Fire Protection Program Document Pkg Section(s): None e.
Code of Federal Regulations Section(s): None f.
Regulatory Guides /NUREGs: 1.97 g.
Other: MasterEauinment List 5.
Describe how the change will affect plant operation when the changed structure, system or component function as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes. Include a discussion of any changed interactions with other structures, systems or components.
During the completion of this portion of the installation work, the service water radiation monitor (SWRM) will be out of service and grab samples will be drawn and analyzed every twelve hours in accordance with technical specification table 3.2-5. Potential effected systems include process radiation monitoring, service water, and domestic water. The process radiation monitoring system will only be affected at the SWRM. These effects are I
described above. The service water system will not be affected by this installation due to the SWRM system being OOS during the described installation. During SWRM system testing, the eductor will draw service water sample flow through the SWRM system and discharge back to the service water return header. This will result in rio change from the present analyzed condition of the system. Domestic water will be used to drive the eductor. The flow will be stopped, started, and throttled, but no adverse effect will be made to the system.
6.
Describe how the change will affect equipment failures. In particular, describe any new failure modes and their impact during all applicable operating modes.
1 A possible failure mode of the Pls would be erratic indication. If this were to occur, system flow set-up could be affected (flow too high or low). Piping for domestic water and service water can withstand the peak system pressure which is domestic water. Possible failure modes of the flow switch is constant no flow and constant normal flow. Each of these conditions is easily detectable and will be detected by performing flow switch setpoint checks.
o:\\CGE\\QAP\\l100\\l100-521 - - - -
CGE QAP 1100-S21 Revision 3 7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident Operation or structure, system or component failure of the changed structure, system or component could lead to the accident ACCIDENT UFSAR SECTION None 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. (To determine the factors affecting the specification, it is necessary to review the UFSAR and SER where the Bases Section of the Technical Specifications does not explicity state the basis).
3.2.G 9.
Will the change involve a Technical Specification revision?
(
) Yes
( X ) No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing step 14, indicate that a Technical Specification revision is required.
G3CGE\\QAM1100 WOO-521.
i CGE QAP 1100-S21 Revision 3 J
10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR may be increased, use one copy of this page to answer the following questions for each accident listed in step 7. Provide the rationale for all NO answers.
Affected accident Not Aonlicable UFSAR Section:
May the probability of the accident be increased?
(
) Yes
(
) No May the consequences of the accident (off-site
(
) Yes
(
) No dose) be increased?
I o;\\CGE\\QAP\\l100\\I100-s21 1
CGE QAP 1100-S21 Revision 3 May the probability of a malfunction of equipment
(
) Yes
(
) No important to safety increase?
May the consequences of a malfunction of equipment
(
) Yes
(
) No important to safety increase?
4 l
l l
If any answer to Ouestion 10 is YES. then an Unreviewed Safety Ouestion exists.
G:\\CGDQAP\\l100\\l100421 5-
i CGE QAP 1100-S21 Revision 3 11.
Based on your answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR7
(
) Yes
( X ) No Describe the rationale for your answer.
'Ihe answer to question 5 details the installation will have no effect on the three interconnecting systems assuming no equipment failures. The answer to question six explored the possible equipment failures and found no adverse impact on the three potentially affected systems. Installation and testing of this equipment cannot cause any plant accident or transient not described within the UFSAR. The installation does not alter the interconnecting systems so as to create abnormal lineups or operating modes. The installation will be passive with respect to the potential to initiate a different type of acciden'.
6 t
If the answer to Ouestion 11 is Yes. then an Unreviewed Safety Ouestion exists.
onCGDQAM1100\\l100-s21 !
CGE QAP 1100-S21 Revision 3 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in step 8. If no Tecimical Specifications are impacted, then no reduction in margin of safety exists, proceed to step 14.
lical Specification 3.2.G l>w..aine which of the following is true for the above specification:
(
)
All caanges to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the rWal acceptance limit need not be identified to determine that no reduction in
' of safety exists, proceed to questio,n 13.
w.
( )
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s)/ margin (s) below.
( )
The applicable parameter or condition change is in a potentially non-conservative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request
'uclear Licensing assistance to identify the acceptance limit / margin for the argin of Safety determination. List the limit (s)/ margin (s) below.
(X)
The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety - proceed to Question 14.
List Acceptance Limit (s)/ Margin (s) of Safety r
13.
Use the above limits identified in step 12 to determine if the margin of safety is reduced (i.e.,
the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
t 0:\\CoE\\QAPil100\\l100-521 7
l CGE QAP 1100-S21 Revision 3 If a Marcin of Safety is reduced. an Unreviewed Safety Ouestion exists.
14.
Check one of the following:
(
)
An Unreviewed Safety Question was identified in step 10, step 11, or step 13.
The proposed change MUST NOT be implemented without NRC approval.
(X)
No Unreviewed Safety Question will result (steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
(
)
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required. Mark below as applicable".
(
)
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
(
)
The change is a plant modification or minor plant change. Mark below as applicable.
(
)
A revision to an existing Technical Specification is required.
The change MUST NOT be installed until receipt of an approved Technical Specification revision.
(
)
The change will not conflict with any existmg Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.
(
)
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation ordy.
Preparer 4(
0 01 June 10.1993 Signature G
Date 03CoE\\QAPtl100\\l100-s21 8
~
CGE QAP 1100-S21 Revision 3 15.
The reviewer has determined that the documentation is adequate to support the above conclusion and agrees with the conclusion.
Reviewer 8MM#
8 Signature Date 16.
Obtain a safety evaluation number from the Tech Staff clerk and record it on page 1.
17.
Leave one copy of the safety evaluation with the Tech Staff clerk and file the original with the applicable package (s) 18.
The Tech Staff clerk will forward a copy of this safety evaluation to the FSAR Coordinator.
(ANI Audit Recommendation 88-1)
Completed:
MC b-ll-3 3 Initial Date (final) o:\\CoEiQAP\\l100\\1100-521
-9
QCAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANS g
L.
Ousd Cities Nuciesr Power Station Reference Number:
$6 //Z Date: f-/ 7-7 9' Sut'=-7 M, /;L. 4,,,
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Submitted by:
lbri,/ M e,, u aw IFOR: REVIEW:snwv.mm-a m.F., vOhW" l
Safety Evaluations.lE irrvoMng an unroviewed safety question as defined in 10CFR50.59 1.
for:
t Changes to procedures as described in the Salery Analysis Report.
s.
Changes to equipment or systems as described in the Safety Analysis Report.
b.
l Tests or --T
..;&,tg described in the Safety Analysis Report.
l c.
Proposed changes which involve an unnsviewed safety question as defined in 10CFM50.59.
_ 2.
a.
)
b.
Equipmera,or sysasen changes..,
j Tests or sixperiments, l
c.
e Proposed changes to the Technical Am or Operating License.
3.
Noncompliance wth codes, regidations, orders, Technical RM license 4.
requirements, or intemel procedures or instructions having nuedaar asisty significance.
Signuicent operating abnormenties or deviations from normal and expected performa 5.
plant egulpmerit that aNoces nuclemr asisty.
8.
M REPORTABLE EVENTS (LERs onlV).
M recognhed indications of an unanticipseM hsf n design or up e, of ". f-i 7.
reisted senatures, systems, or components. -
M changes to the Station Emergency Plan prks ta ime 8.
M ltems referred by the Systems Engineming Supervisor, Station Manager, Ste Vice 9.
President, and Genomi Manager of Oussty Programs and Assessments.
vm --
'g MD"41@m ow-<;~,,
..a.,,
. & ~,, L*biWFW3.WdWfM"w p <**myr yewswwmmuunwenw IFDR1NFORMATION:
- 10. Other OSR lesms/DocumerasJE addressed above.
This Transmutal is being made in accontence wth Quad Chios Nuclear Power Station Technical 5-- r-- =-e 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and investigative Punctiort l
i 8
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CGE QAP 1100-521 Revision 3 10CFR50.59 SAFETY EVALUATIONS January 1993 Safety Evaluation Number: SE D -ll 1 Document Identifier: M04-lO)-89-115 work nackaces 002824. 002825 mechanical scope (Modification, Temp Alt, Work Request Number, etc.)
Unit (s): 10)
System (s):
1700.3900.4200 Applicable Plant Mode (s): This evaluation is aonlicable for all niant operatine modes.
(RUN, STARTUP/ HOT STNBY, REFUEL or SHUTDOWN)
Plant Mode Restriction (s): This evaluation contains no mode restrictions.
1.
Describe the proposed change:
The work to be performed under these packages will demolish the existing service water radiation monitor (SWRM) sample delivery system (receiver tank, pump, all associated piping and valves) and install a new, eductor driven system powered by domestic water. The sample system inlet isolation valve will be replaced and the service water return header will be open through a 1*-1/2" pipe to the turbine building 595* level during the replacement. Once installed, this valve will act as the isolation point for further installation work. Domestic water will be isolated for installation of the back flow preventer. All other items (skid, detector, eductor) will then be installed. A flow indicator will be installed to facilitate Instrument Maintenance work and testing on the flow switch and pressure gauges. The indicator will be removed and replaced with the switch. A leak test will then be performed.
2.
Reason for the change:
This work package was written to install the mechanical portion of rnodification M04-1(2) 115.
3.
Is the change:
( X ) Permanent
(
) Temporary - Expected Duration:
OnCoE\\QAP\\ll00\\l100-s21
-].
CGE QAP 1100-S21 Revision 3 4.
List the reference documents reviewed which describes the structure, system or component.
(Identify documents referenced even if no information was found in that section.)
a.
UFSAR Section(s): Table 1.8-1. 3.0. 7.1. 7.5.11.5.2.7.15 b.
SER Section(s): _None c.
Tech Spec Section(s): 3.2/4.2. 3.8/4.8 d.
Fire Protection Program Document Pkg Section(s): FPR Vol. 14.7.3.4.7.4 e.
Code of Federal Regulations Section(s): None f.
Regulatory Guides /NUREGs: 1.97 g.
Other: SE 93-109 5.
Describe how the change will affect plant operation when the changed structure, system or component function as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes. Include a discussion of any changed interactions with other structures, systems or components.
During installation of this modification, continuous monitoring of the service water return header will be lost. Instead, the Chemistry Department will take and analyze grab samples in accordance with the requirements of Technical Specification Table 3.2-5. Domestic water will be isolated to the Fish House and Chemistry Labs during installation of the back flow preventer, but this will not affect any system needed to assure continued normal plant operations. During replacement of the inlet isolation valves, the intake sample line will be open to turbine building atmosphere. A funnel and drain hose will be in place to collect any water and route it to a floor drain. Some water may not be caught by the funnel, but the small i
diameter of the pipe (l'-1/2") means the volume can easily be contained by local floor drains.
i No other structures, systems, or components (SSC) will be affected by this work.
6.
Describe how the change will affect equipment failures. In particular, describe any new failure modes and their impact during all applicable operating modes.
I The new isolation valve will be a ball valve. The old isolation valve was a gate valve which was susceptible to stem / disk separation. A possible failure mode would be the ball valve sticking partially or fully open after it is installed. This would result in leakage of service water onto the floor. Nearby floor drains, however, will be able to handle this leakage and l
prevent local flooding. This analysis is also good for any leakage after the sample supply system. The piping supports could fail resulting in an improperly supported line. This may or may not lead to line cracking or rupture, but the resulting leakage is still bounded by the above analysis. This work will not affect any other systems or impact the present failure modes of any SSC.
0:\\CoE\\QAPil100\\ll04521
-2
CGE 1
QAP 1100-S21 Revision 3 7.
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis The changed structure, system or component is explicitly or implicitly assumed to i
function during or after the accident Operation or structure, system or component failure of the changed structure, system or component could lead to the accident ACCIDENT UFSAR SECTION None 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. (To determine the factors affecting the specification, it is necessary to review the UFSAR and SER where the Bases Section of the Technical Specifications does not explicitly state the basis).
3.2.G 9.
Will the change involve a Technical Specification revision?
(
) Yes
( X ) No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing step 14, indicate that a Technical Specification revision is required.
0:\\CoE\\QAP\\l100\\l100-521 CGE QAP 1100-S21 Revision 3 10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR may be increased, use one j
copy of this page to answer the following questions for each accident listed in step 7. Provide the rationale for all NO answers.
Affected accident Not Applicable UFSAR Section:
May the probability of the accident be increased?
(
) Yes
(
) No May the consequences of the accident (off-site
(
) Yes
(
) No dose) be increased?
C i
0:\\CGE\\QAP\\l100\\l104S21 CGE QAP 1100-S21 Revision 3 May the probability of a malfunction of equipment
(
) Yes
(
) No important to safety increase?
i l
May the consequences of a malfunction of equipment
(
) Yes
(
) No important to safety increase?
l r
If any answer to Ouestion 10 is YES. then an Unreviewed Safety Ouestion exists.
l 0:\\CGE\\QAP\\l100\\l100-521. -
CGE QAP 1100-S21 Revision 3 11.
Based on your answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR?
(
) Yes
( X ) No Describe the rationale for your answer.
This work only interfaces with the domestic water system and the service water system. Both interfaces are mechanical only. No other SSC will be impacted by the scope of this work. The worst case scenario would involve a failure of the installed isolation valve on the service water return header. This would lead to leakage onto the turbine building first floor. But, this leakage will not be of greater magnitude than the capability to remove water by the floor drain system. Therefore, this event will not result in flooding. No other SSCs will be adversely impacted so as to create a new UFSAR accident or transient.
t l
If the answer to Ouestion 11 is Yes. then an Unreviewed Safety Ouestion exists.
o:\\CoEiQAPil100\\l100-321
.(y.
CGE QAP 1100-S21 Revision 3 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in step 8. If no Technical Specifications are impacted, then no reduction in margin of safety exists, proceed to step 14.
Technical Specification 3.2.G Determine which of the following is true for the above specification:
(
)
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists, proceed to question 13.
( )
The Technical Specification provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s)/ margin (s) below.
( )
The applicable parameter or condition change is in a potentially non-conservative direction and the Technical Specification neither provides an acceptance limit nor explicitly references a limit in the UFSAR. Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination. List the limit (s)/ margin (s) below.
(X)
The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety - proceed to Question 14.
t List Acceptance Limit (s)/ Margin (s) of Safety i
13.
Use the above limits identified in step 12 to determine if the margin of safety is reduced (i.e.,
the new values exceed the acceptance limits). Describe the rationale for your determination, i
Include a description of compensating factors used to reach that conclusion.
1 If a Marcin of Safety is reduced. an Unreviewed Safety Ouestion exists.
O ACoBQAM1100u l00-521 7
CGE QAP 1100-S21 Revision 3 14.
Check one of the following:
(
)
An Unreviewed Safety Question was identified in step 10, step 11, or step 13.
The proposed change MUST NOT be implemented without NRC approval.
(X)
No Unreviewed Safety Question will result (steps 10,11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
(
)
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is regnired. Mark below as applicable.
(
)
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
(
)
The change is a plant modification or minor plant change. Mark below as applicable.
(
)
A revision to an existing Technical Specification is required.
The change MUST NOT be installed until receipt of an approved Technical Specification revision.
(
)
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.
(
)
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only.
Preparer N-June 15.1993 Signattire
[~
Date 0;\\CoE\\QAP\\1100\\1100421 8
t
CGE QAP 1100-S21 Revision 3 15.
The reviewer has determined that the documentation is adequate to support the above conclusion and agrees with the conclusion.
ReviewerMMM 4
Signature Date 16.
Obtain a safety evaluation number from the Tech Staff clerk and record it on page 1.
17.
Leave one copy of the safety evaluation with the Tech Staff clerk and file the original with the applicable package (s) 18.
The Tech Staff clerk will forward a copy of this safety evaluation to the FSAR Coordinator.
(ANI Audit Recommendation 88-1)
Completed:
hK bo-M-N Initial Date P
i i
(final) 0:\\CoE\\QAP\\1100\\1100-521 9
l
.u QCAP 1000-6 UNIT 1(2)
REVISION 0 ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Ouad Cities Nuclear Power Station
- h Y Coh- /. 9 7 - o 9M Date:
Reference Number:
Subject:
84dA d 6 7Z4FIpMC h4 WWAc bMC, M 5:cTL"4Eit
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Submitted by:
d f, d di5 T'E d FOR REVIEW:
1.
Safety Evaluations NOT involving an unreviewed safety question as defined in g
10CFR50.59 for:
l a.
Changes to procedures as described in the Safety Analysis Report.
L b.
Changes to equipment or systems as described in the Safety Analysis f
- Report, c.
Testa or experiments NOT described in the Safety Analysis Report.
I 2.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59.
i a.
Procedure changes.
b.
Equipment or system changes.
c.
Tests or experiments.
j 3.
Proposed changes to the Technical Specifications or Operating License.
i 4.
Noncompliance with codes, regulations, orders, Technical Specifications, license i
requirements, or internal procedures or instructions having nuclear safety significance.
l 5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
All REPORTABLE EVENTS (LERs only).
7.
All recognized indications of an unanticipated deficiency in design or operation of safety related structures, systems, or components.
l i
8.
All changes to the Station Emergency Plan prior to implementation.
9.
All items referred by the Systems Engineering Supervisor, Station Manager, Site Vice
}
President, and General Manager of Quality Programs and Assessments.
i FOR INFORMATION:
y'
- 10. Other OSR ltems/ Documents NOT addressed above.
T This Transmittal'is being made in accordance with Quad Cities Nuclear Power Station Technical Specifications 6.1 G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and investigative Function.
4 4
8
Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 1 of 10 Station / Unit Ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
Enaineerina Chance Notice 04-01031E dated 12/6/93.
Bechtel Calculation OC-429-C-035 dated 12/7/93.
2.
Describe the proposed change and the reason for the change.
The subject exempt change will replace an oil-filled 1 MVA transformer with a dry-type 500 KVA transformer on elevation 639' of the Unit 1 Turbine Building.
The existing wet pipe system will be demolished.
3.
Is the change:
[X]
Permanent
[]
Temporary -
Expected duration AND Plant Mode (s) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected syste.ns, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation. List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
The following FSAR sections were reviewed for applicability:
8.0 Electric Power 8.3 Onsite Power Systems The change does not affect these documents.
QE-06.1 DECA Version 2.3
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Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 2 of 10 Station / Unit ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
T42R-5A receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kv system is not used for plant equipment. Therefore, this transformer will not electrically affect operation of plant equipment.
Per the Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
6.
Describe how the change will affect equipment failures.
In particular, describe any new failure modes and their impact during all applicable operating modes.
i This change does not electically or structurally interact with plant equipment. Therefore, equipment failures are not affected.
The failure mode of the transformer is not changed.
7.
Identify each accident or anticipated transient (i.e.,
large/small break LOCA, loss of load, turbine missiles, fire, flooding) described in the SAR where any of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or failure of the changed SSC could lead to the accident I
ACCIDENT SAR SECTION None.
8.
List each Technical Specification (Safety Limit, Limiting Safety System i
Setting or Limiting Condition for Operation) where the requirement, l
associated action items,. associated surveillances, or bases may be affected. To determine the factors affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.
t None.
i
)
QE-06.1 DECA Version 2.3
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4 Exhibit E Mod # K04-1-93-094 ENC-QE-06.1 l
Revision 5 Page 3 of 10 Station / Unit Ouad Cities
/1 Exhibit E
.j 10CFR50.59 SAFETY EVALUATION 9.
. Will the change involve a Technical Specification revision?
[] Yes
[X] No If a Technical Specification revision is involved, the change cannot be o
implemented until the NRC issues a license amendment. When completing Step 14, indicate that a Technical Specification revision is required.
i i
QE-06.1 DECA Version 2.3
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Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5
)
Page 4 of 10 Station / Unit Ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident licted in Step 7.
Provide the rationale for all NO answers.
Af fected accident None.
SAR Section:
N/A.
May the probability of the accident be increased?
[] Yes
[X] No T42R-5A receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kv system is not used for plant equipment. Therefore, this transformer will not electrically affect operation of plant equipment.
Per the Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
The probability of an oil fire due to transformer failure is reduced to zero because the new transformer is a dry-type containing no oil.
May the consequences of the accident (off-site dose)
[] Yes
[X] No be increased?
T42R-5A receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kv system is not used for plant equipment. Therefore, this transformer will not electrically affect operation of plant equipment.
Per the Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
Therefore, the transformer has no affect on the consequences of an accident.
The consequences of failure of the transformer is reduced significantly since the dry-type transformer will not contribute combustible oil to a fire in the immediate area.
May the probability of a malfunction of equipment
[] Yes
[X] No important to safety increase?
T42R-5A receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kv system is not used for plant equipment. Therefore, this transformer will not QE-06.1 DECA Version 2.3
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[
f Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 5 of 10 Station / Unit Duad cities
/1 i
Exhibit I 10CFR50.59 SAFETY EVALUATION electrically affect operation of plant equipment.
Per the Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
The replacement of the oil-type transformer with a dry-type one i
results-in no net change in probability of malfunction of safety related equipment.
May the consequences of a malfunction of equipment
[] Yes
[X] No important to safety increase?
T42R-5A receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kV system is not used for l
plant equipment. Therefore, this transformer will not electrically affect operation of plant equipment.
Per the i
Bechtel calculation listed previously, the supports and attachments have been evaluated for structural acceptability.
l i
The replacement of the oil-type transformer with a dry-type one results in no net change in consequences.of malfunction of safety related equipment.
The dry type transformer will reduce the consequences of transformer failure or area fire.
i If any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
f l
t
)
I l
QE-06.1 DECA VeTsion 2.3 s
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l J
Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 6 of 10 Station / Unit Ouad Cities
/1 l
i Exhibit E 10CFR50.59 SAFETY EVALUATION t
i 11.
Based on your answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the l
SAR?
.i
(] Yes (X) No l
1 Describe the rationale for your answer.
l T42R-SA receives power from the 13.8KV yard.
Its 480V secondary will provide power for maintenance activities on the turbine deck.
Per the FSAR, the 13.8kV system is not used for plant equipment. Therefore, this transformer will not electrically affect operation of plant equipment.
Per the Bechtel calculation' listed previously, the supports and i
attachments have been evaluated for structural acceptability.
The replacement of the oil-type transformer with a dry-type one l
results in no new accident type.
l I
If the answer to Question 11 is Yes. then an Unreviewed Safety Question exists.
I i
I t
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i i
i l
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i 1
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Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 7 of 10 Station / Unit Ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
N/A.
(Check appropriate condition) :
() All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.
[]
The Technical Specification or SAR provides a margin of safety e acceptance limit for the applicable parameter or condition. List the limit (s) / margin (s) and applicable reference for the margin of safety below - proceed to question 13.
l
[]
The applicable parameter or condition change is in a potentially i
non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit.
Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references.
List the agreed limit (s) / margin (s) below.
t
[X]
The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety.
Proceed to question 14.
]
List Acceptance Limit (s) / Margin (s) of Safety i
Tech Spec 1
SAR Section SER Section QE-06.1 DECA Version 2.3
i Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 l
Revision 5 j
Page 8 of 10 Station / Unit Ouad Cities
/1 i
Exhibit E 10CFR50.59 SAFETY EVALUATION
' 13.
Use the above limits to determine if the margin of safety is reduced i
(i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion, j
f i
i If a Marain of Safety is reduced an Unreviewed Safety Question existh i
i P
t i
I i
t i
l l
I L
I QE 06.1 DECA Version 2.3
m Exhibit E Mod # E04-1-93-094 ENC-QE-06.1 Revision 5 Page 9 of 10 Station / Unit Ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 14.
Check one of the following:
[}
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
[:X]
No Unreviewed Safety Question will result ( Steps 10, 11, and 13)
AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
[]
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.
Mark below as applicable.
[]
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
[]
The change is a plant modification or minor plant change.
Mark below as applicable.
[]
A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt of an approved Technical Specification revision.
[]
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required.
In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.
[}
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only.
i f
QE-06.1 DECA Version 2.3
y.
t Exhibit E
- Mod # E04-1-93-094 ENC-QE-06.1.
Revision 5 Page 10 of 10 Station / Unit Ouad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION Note:
Par ial Modi ications and/or separate 10CFR50.59 reviews for or' ion of e work may be used to facilitat installation.
Preparer (Cogn ant Engineer)
Date 15.
The rev' ewer has determined that the documentation is adequate to support the above conclusion and agrees with the conclusion.
Revievet k
m or m A
/- A /- 9 Y (Design Superintendent / Supervisor)
Date QE-06.1 DECA Version 2.3 1.
QCAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station 6DM-/- 9 '/ - hf 4[E a 4 Date:
Reference Number:
Subiect:
/Mf'kL L tJtLPsd,
act ccPT40.LCS od
'77x /E 6 & G d
9hiGLP N NL L-Subrnitted by:
J.,% S N FOR REVIEW:
1.
Safety Evaluations NOT involving an unreviewed safety question as defined in 10CFR50.59 for:
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
c.
Tests or experiments NOT described in the Safety Analysis Report.
2.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59, a.
Procedure changes.
b.
Equipment or system changes.
c.
Tests or experiments.
3.
Proposed changes to the Technical Specifications or Operating License.
4.
Noncompliance with codes, regulations, orders, Technical Specifications, license requirements, or internal procedures or instructions having nuclear safety significance.
5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
All REPORTABLE EVENTS (LERs only).
7.
All recognized indications of an unanticipated deficiency in design or operation of safety-related structures, systems, or components.
8.
All changes to the Station Emergency Plan prior to implementation.
9.
Allitems referred by the Systems Engineering Supervisor, Station Manager, Site Vice President, and General Manager of Quality Programs and Assessments.
FOR INFORMATION:
K
- 10. Other OSR ltems/ Documents NOT addressed above.
This Transmittalis being made in accordance with Quad Cities Nuclear Power Station Technical Specifications 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and Investigative Function.
8
Mod 8 E04-1-93-245 Exhibit E ENC-QE-06.1 Page 1 of 8 Station / Unit Quad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
Exempt Plant Change E04-1-93-245.
Engineering Change Notice 04-10532, dated 1/17/94.
Bechtel Calculation QC-469-C-001, dated 1/21/94.
Parameter Assessment and Reconciliation B-130-00360.
2.
Describe the proposed change and the reason for the change.
The subject exempt change will replace the existing panel with a 10 circuit distribution panel and install six 60 amp welding receptacles powered from this new pane].
Five receptacles will be mounted on the outside of the turbine shield wall west of the new circuit panel. A sixth will be installed on the inside of the turbine shield wall.
This new configuration will provide a safer and more officient means for providing power on the turbine deck.
3.
Is the change:
[X]
Pemanent
[}
Temporary -
Expected duration AND Plant Mode (s) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which discuss the affected SSCs or their operation. List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
The following UFSAR Sections were reviewed:
1.2.2.10 Shielding, Access Control, and Radiation Protection Procedures 3.5 Missile Protection 8.0 Electric Power Systems 8.3 Onsite Power Systems 10.0 Steam and Power Conversion System These sections will not be affected by this design change.
5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i e.,
focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
Include a discussion of any changed interactions with other SSCs.
' The power source for the new 480VAC panel and welding receptacles is 1
from the 13.BKV yard via transformer T42R-5A.
Per the UFSAR, the 13.8KV yard is not used for plant equipment.
Therefore, the subject design change will not ele::;rically affect plant equipment.
The only structural interaction la due to mounting on the turbine shield wall (which is non-safety related / non-ceismic / non-II/I).
The referenced calculation and associated PAR have detemined that the stinctural loads are acceptable.
QE-06.1 DECA Version 2.3 i
.. _ ~
Mod 8 304-1 93-245 Exhibit E ENC-Q3-06.1 Station / Unit Quad Cities
/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 1
6.
Describe how the change will affect equipment failures.
In particular, i
describe any new failure modes and their impact during all applicable i
operating modes.
As stated in Section 5, the design change does not electrically interact with plant equipment.
The additional structural loads have been analyzed for acceptability.
Therefore; the design change will not affect equipment failures.
The addition of the circuit panel and welding receptacles will result in an increase in equipment.
reliability over the existing configuration.
Therefore, the failure mode of the Af 0VAC panel is lessened in severity.
7.
Identify each accident or anticipated transient (i.e., large/small break LOCA,.
loss of load, turbine missiles, fire, flooding) described in the SAR where any of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or failure of the changed SSC could lead to the accident ACCIDENT SAR SECTION None N/A 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where.the requirement, associated action items, associated surveillances, or bases may be affected. To
~
determine the factors affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitly state the basis.
The following sections were reviewed:
3.9/4.9 Auxiliary Electric Systems 9.
Will the change involve a Tcchnical Specification revision?
[] Yes
[X] No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing Step 14, indicate that a Technical Specification revision _is required.
i i
l
- i i
)
QE-06.1 DECA Version 2.3 i
Mod i C04-1-93-245 Exhibit E ENC-QE-06.1 Station / Unit Quad Cities
/1 Exhibit I 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7 Provide the rationale for all NO answers.
Affected accident: None SAR Section:
N/A May the probability of the accident be increased?
[] Yes
[X] No As stated in section 5, the subject exempt change does not electrically interact with plant equipment.
The structural interactions are with plant struccares that are not important to safety and have been analyzed for acceptability.
The replacement circuit panel and receptacles will be more reliable than the existing configura tion.
Therefore, the chances of failure of this equipment is reduced.
The probability oE an design basis accident is unchanged.
May the consequences of the accident (off-site dose)
[] Yes
[X] No be increased?
As stated prev. usly, the subject design change does not affect equipment important to safety or equipment required for safe shutdown of the plant.
Therefore, the consequences of an accident are not changed by this change.
May the probability of a malfunction of equipment
[] Yes
[X] No important to safety increase?
As previously stated, the subject exempt change does not electrically interact with plant equipt"en t.
The structural interactions are with plant structures that are not important to safety and have been analyzed for acceptability.
The replacement circuit panel and receptacles will be more reliable than the existing configuration.
Therefore, the chances of failure of this equipment is reduced.
Therefore, the probability of calfunction of other nearby equipment due to failure of this equipment is reduced.
May the consequences of a malfunction of equipment
[] Yes
[X) No important to safety increase?
As previously stated, the subject exempt change does not electrically interact with plant equipment.
The structural interactions are with plant structures that are not importan t to cafety and have been analyzed for acceptability.
The replacement circuit panel and receptacles will be more reliable than the existing configuration.
The failure mode of this new panel is the same as the existing panel.
Such a failure does not change the consequences of a calfunction of
\\
other nearby equipment.
Therefore, it will not change the consequences of safety related equipment failure.
j If any answer to Question 10 is YES. then an Unreviewed Safety Question exists a
)
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Exhibit E 10CFR50.59 SAFETY EVALUATION 11.
Based on your answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the SAR?
l t
[] Yes (X) No Describe the rationale for your answer.
The subject des:gn change will not result in changed operation of the existing panels.
Therefore, no new accident that has not been previously analyzed will be created.
i If the answer to Question 11 is Yes. then an Unreviewid Safety Question exists.
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/1 Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following Questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation.
N/A Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
N/A (Check appropriate condition):
[]
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction.
Therefore, the actual acceptance limit need not be identified to i
determine that no reduction in margin of safety exists - proceed to I
Question 13.
[]
The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s) / margin (s) and applicable referencu for the margin of safety below - proceed to question 13.
l
[]
The applicable parameter or condition change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit.
Request Nuclear Licensing assistance to identify the acceptance j
limit / margin for the Margin of Safety determination by consulting the i
NRC, SAR, SER's or other appropriate references.
List the agreed i
limit (s) / margin (s) below.
[X)
The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety.
Proceed to question 14.
List Acceptance Limit (s) / Margin (s) of Safety Tech Spec l
t SAR Section l
l SER Section j
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10CFR50.59 SAFETY EVALUATION i
13.
Use the above limits to determine if the margin of safety is reduced (i.e.,
the new values exceed the acceptance limits). Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
N/A l
If a Marain of Safety is reduced an Unreviewgd Safety Question exists.
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/1 i
Exhibit I 10CFR50.59 SAFETY EVALUATION 14.
Check one of the following:
[]
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
[X]
No Unreviewed Safety Questien will result ( Steps 10, 11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
[]
A Technical Specification revision is involved; but no Unreviewed Safety l
Question will result. The proposed change requires a License Amendment.
Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.
Mark below as applicable, f
()
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59.
Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
[]
The change is a plant modification or minor plant change. Mark i
below as applicable.
[]
A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt.of an approved Technical Specification revision.
[]
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required.
In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization is granted, the block below should be checked.
[]
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only.
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/1 Exhibit E 10CFR50.59 SAFETY EVALUATION Note: P ti 1 die ications and/or separate 10CFR50.59 reviews for portions of h w rk y beusedtofacilitateinstalapion.
N(
' t /* 'l Preparer (C
izant Engineer)
Date 15.
The re';iewer has determined that the documentation is adequate to support the above conclusion and agrees with the conclusion.
Reviewer
- h M'I' Y (Design Superintendent / Supervisor)
Date f
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L QE-06.1 DECA Version 2.3 t
QCAP 1000-6 UNIT 1(2)
REVISION O ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station j
k9Y Reference Number:
CD - / - 9 ~5/- 97 W Date:
t Subiect:
Crkf WCgC-n r C & GL% isc" t:.3 dfUc.-
i Submitted by:
M @. A.lM FOR REVIEW:
1.
Safety Evaluations NOT involving an unreviewed safety question as defined in 10CFR50.59 for:
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equipment or systems as described in the Safety Analysis Report.
c.
Tests or experiments NOT described in the Safety Analysis Report.
2.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59.
i i
a.
Procedure changes.
l b.
Equipment or system changes.
j c.
Tests or experiments.
3.
Proposed changes to the Technical Specifications or Operating License.
4.
Noncompliance with codes, regulations, orders, Technical Specifications, license requirements, or internal procedures or instructions having nuclear safety significance.
5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
All REPORTABLE EVENTS (LERs only).
7.
All recognized indications of an unanticipated deficiency in design or operation of safety-related structures, systems, or components.
8.
All changes to the Station Emerger'cy Plan prior to implementation.
9.
Allitems referred by the Systems Engineering Supervisor, Station Manager, Site Vice President, and General Manager of Quality Programs and Assessments FOR INFORMATION:
[
- 10. Other OSR ltems/ Documents,N_QT addressed above.
This Transmittalis being made in accordance with Quad Cities Nuclear Power Station Technical Specifications 6.1.G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and Investigative Function.
8 4
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5 nod n E04-1-93-325 Exhibit E ENC-QE-06.1 i
Page 1 of 8 Station / Unit Ouad Cities
/ 1 Exhibit I 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing the proposed change.
Exempt Plant Change E04-1-93 325.
Engineering Change Notice 04-001149S and associated calculations.
2.
Describe the proposed change and the reason for the change.
The subject exempt plant change will install new concrete piers in support of the replacement of the Unit 1 Unit Auxiliary Transformer (UAT). New concrete piers are required for the fire suppression deluge system which is being redesigned due to physical differences between the existing GE and the new SMIT transformer.
Two other exempt changes are required to complete the replacement of the Uni:
1 UAT:
E04-1-93-326 will replace the existing fire protection system piping and fire detection method.
The deluge piping must be replaced due to the physical differences between the existing GE UAT and the new SMIT UAT.
The detection method is being changed in order to make it more reliable. The overall operation of the system will not change.
E04-1-93-327 will reinstall the transformer control circuitry.
These changes are necessazy due to slight diiferences between the GE and SMIT transformers.
The control circuitry changes will not atfect the operation of the plant.
3.
Is the change:
(X]
Permanent
[]
Temporary -
Expected duration AND Plant Mode (s) restrictions while installed (NONE if no plant mode restrictions apply) 4.
List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis secticr.s which discuss the affected SSCs or their operation.
List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.
The following section of the Quad Cities Station UFSAR has been reviewed Section 8.3 Onsite Power Systems 5.
Describe how the change will affect plant operation when the changed SSCs function as intended (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
1 Include a discussion of any changed interactions with other SSCs.
The replacement of the Unit 1 General Electric UAT with a new SMIT UAT does not affect plant operation in any operating mode.
The new UAT has electrical characteristics that are compatible with the existing GE UAT.
It has the QE-06.1 DECA version 2.3 i
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. l Nod S E04-1-93-325 Cxhibit E ENC-QE-06.1 Page 2 of 8 Station / Unit Duad Cities
/ 1 f
I Exhibit E 10CFR50.59 SAFETY EVALUATION capability to supply auxiliary power for the unit while in run mode or in backfeed mode.
The control wiring changes are necessary due to the new UAT's l
slight differences and enhancements.
The replacement of the fire protection deluge piping and installation of associated concrete supports are necessary due to the minor differences in the new VAT's physical layout.
The replacement of the existing thermal detectors with linear-type detection cable j
will improve the reliability of the system by reducing the inadvertent actuations. These changes to the Unit 1 UAT do not change operation of the UAT as it relates to the plant or plant systems.
6.
Describe how the change will affect equipment failures.
In particular, describe any new failure modes and their impact during all applicable operating modes.
The failure mode of the new UAT is the same as the existing UAT.
The failure i
E rate of the new UAT should be lower than that of the existing UAT due to the age of the existing UAT.
The replacement of the deluge system piping and installation of associated concrete pad supports does not affect any other equipment.
The failure mode j
of these components is the same as for the existing systent.
The change in fire detection method will increase the fire detection l
reliability.
The replacement "protecto-wire" detection method will reduce the l
number of DC ground problems, and therefore increase the reliability of the e
system.
The failure mode of the fire detection system is the same as for the existing detectors.
7.
Identify each accident or anticipated transient (i.e.,
large/small break LOCA,
}
loss of load, turbine missiles, fire, flooding) described in the SAR where any i
l of the following is true:
The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function'during 2
cc after the accident Operation or f ailure of the changed SSC could lead to the accident ACCIDENI SAR SECTION Loss of Auxiliary Power 8.3.1 8.
List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine the factets affecting the specification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitly state the basis.
The following section was reviewed:
Section 3.9.A.3, "One other 345-kv line capable of carrying auxiliary power to an essential electrical bus of the unit through the 4160-volt bus tie shall '
be available."
QE-06.1 DECA Version 2.1
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/ 1 Exhibit I 10CFR50.59 SAFETY EVALUATION 9.
Will the change involve a Technical Specification revision?
[] Yes (X) No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When completing' Step 14, indicate that a Technical Specification revision is required.
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Hod 0 E04-1-93-325 Exhibit E ENC-QE-06.1 Page 4 of 8 Station / Unit-Ouad Cities
/ 1 Exhibit I 10CFR50.59 SAFETY EVALUATION 10.
To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each accident listed in Step 7.
Provide the rationale for all NO answers.
Affected accident: Loss of Auxiliary Power SAR Section:
Section 8.3.1 May the probability of the accident be increased?
[] Yes
[X] No The replacement UAT will provide the same function as the existing UAT.
The new UAT should be more reliable than the existing UAT because it is new and has been constructed using latest technology.
Therefore, the probability of '
the accident will not increase by this change.
May the consequences of the accident (off-site dose)
[] Yes (X) No be increased?
The consequences on plant operation of failure of the new UAT are the same as for the old UAT.
Therefore, the consequences of a failure of the Unit 1 UAT will not change due to the replacement of the existing GE UAT with the new SMIT UAT.
May the probability of a malfunction of equipment
[] Yes (X) No important to safety increase?
This change is compatible with interfacing plant systems.
The replacement of the GE UAT with a new SMIT UAT will reduce the probability of a Unit 1 UAT failure by improving the reliability of the transformer and fire detection circuitry.
Therefore, the probability of a malfunction of equipment important i
to safety will be reduced.
May the consequences of a malfunction of equipment
[ ] Yes (X) No important to safety increase?
The consequences of failure of the UAT or the associated fire protection system is the same as for the existing UAT.
Therefore, the consequences of failure of equipment important to safety is unchanged as a result of this project.
If any answer to Question 10 is YES, then an Unreviewed Safety Question exists.
t l
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Exhibit E nod i E04-1-91-125 ENC-QE 06.1 Page 5 of 8 Station / Unit ouad Cities
/ 1 Exhibit E 10CFR50.59 SAFETY KVALUATION-
)
11, Based on your answers to Questions 5 and 6, does the change adversely impact systems or functions so as to create the possibility of an accident or i
malfunction of a type different from those evaluated in the SAR7
[] Yes (X) No Describe the rationale for your answer.
The UAT is being replaced by a never transformer.
The failure mode of this new transformer, fire protection system, and control circuitry is the same as i
for the existing transformer.
The failure race due to these changes is reduced due to the more reliable transformer and enhancements to the fire protection system.
Therefore, an accident ditferent from those previously evaluated in the SAR is not created.
t If the answer to Question 11 is Yes. then an Unreviewed Safety Question exists.
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. Hod i E04-1-93-325 ENC-QE-06.1 Page 6 of 8 Station / Unit Ouad Cities
/ 1 Exhibit I 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following Questions for each Technical Specification listed in Step 8.
List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed i
I for this evaluation.
Section 3.9.A 3, "One other 345-kv line capable of carzying auxiliazy power to an essential electrical bus of the unit through the 4160-volt bus tie shall be available. "
Evaluation of Technical Specification (Enter N/A if none are affected and check last option.)
The replacement of the UAT does not affect the requirements for this Technical Specification.
The new UAT will be capable of providing auxiliary power to the unit during backfeeding operations (Check appropriate condition):
i
[]
All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction.
l Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to l
Question 13.
[ ]
The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. List the f
t limit (s) / margin (s) and applicable reference for the margin of safety below - proceed to question 13.
j
[]
The applicable parameter or condition change is in a potentially I
non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit.
Request Nuclear Licensing assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references. List the agreed limit (s) / margin (s) below.
[X)
The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin.
of safety.
Proceed to question 14.
List Acceptance Limit (s) / Margin (s) of Safety Tech Spec SAR Section SER Section i
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ENC-QE-06.1 i
Page 7 of 8 Station / Unit Ouad Cities
/ 1 Exhibit E 10CFRSO.59 SAFITY EVALUATION 13.
Use the above limits to detennine if the margin of safety is reduced (i.e.,
the new values exceed the acceptance limits).
Describe the rationale for your determination.
Include a description of compensating factors used to reach that conclusion.
N/A If a Marain of Safety is reduced an Unreviewed Safety Question exists.
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/ 1 Exhibit E 10CFR50.59 SAFETY EVALUATION r
14.
Check one of the following:
[]
An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13.
The proposed change MUST NOT be implemented without NRC approval.
[X)
No Unreviewed Safety Question will result ( Steps 10, 11, and 13) AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.
[]
A Technical Specification revision is involved; but no Unreviewed Safety Question will result. The proposed change requires a License Amendment.
Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required. Mark below as applicable.
~
[]
The change is not a plant modification or minor plant change and will not be implemented under 10CFR50.59. Upon receipt of the approved Technical Specification change from the NRC, the change may be implemented.
[]
The change is a plant modification or minor plant change.
Mark below as applicable.
[]
A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt cf an approved Technical Specification revision.
[]
The change will not conflict with any existing Technical Specifications and only new Technical Specifications are i
required. In these cases, Nuclear Licensing may authorize i
installation, but not operation, prior to receipt of NRC approval of the License Amendment.
If such authorization :s granted, the block below should be checked.
[]
Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for. installation only.
Note: Partial od ficttions and/or separate 10CFR50.59 revieas for portions cf t
r ma be used to facilitate installation.
D Preparer
~
N (Co izant Engineer)
Date 15.
The r iewer has determined that the documentation is adequate to support the ove conclusion and agrees with the conclusion.
- [/hV Reviewer amm em -
(Design Superintendent / Supervisor)
Datd f
QE-06.1 DECA Version 2.3
_. ~ _
QCAP 1000-6
- ~
~
UNIT 1(2)
REVISION 0 ATTACHMENT A (Page 1 of 1)
OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRANSMITTAL Quad Cities Nuclear Power Station Reference Number: @ 4 - \\ - J \\ - \\'l'l Date: (o[8/94-
Subject:
EZEPLtcEMEOT OF EroETtM % T GX')EcoAGf?_
Eca._T 5 a
A Submitted by: M hAIs[4v v
t FOR REVIEW:
1.
Safety Evaluations.tiQI involving an unreviewed safety question as defined in 10CFR50.59 for:
a.
Changes to procedures as described in the Safety Analysis Report.
b.
Changes to equiprnent or systems as described in the Safety Analysis Report.
c.
Tests or experiments RQI described in the Safety Analysis Report.
2.
Proposed changes which involve an unreviewed safety question as defined in 10CFR50.59.
a.
Procedure changes.
j b.
Equipment or system changes.
c.
Tests or experiments.
3.
Proposed changes to the Technical Specifications or Operating License 4.
Nercvw.pl lance with codes, regulations, orders, Technical Specifications, license requirements, or intomal procedures or instructions having nuclear safety significance.
5.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety.
6.
M REPORTABLE EVENTS (LERs only).
7.
M recognized indications of an unanticipated deficiency in design or operation of safety-reisted structures, systems, or components.
j 8.
M changes to the Station Emergency Plan prior to implementation.
M tems referred by the Systems Erg' :: "is Supervisor, Station Manager, Site Vice l
9.
c-Proeident, and General Manager of Quality Programs and Assessments.
FOR"INFORMATION: #
[
- 10. Other OSR ltems/DocumentsJE addressed above This Transmittal is being made in accordance with Quad Cities Nuclear Power Station Technical Specifications 6.1 G.2.d(1) for information only. No specific action is required unless deemed necessary by Offsite Review and investigative Function 8
Mod o Po4 91-127 Exhibit E ENC OE 06.1 Revision 3 OUAD cit 1Es/l Page 1 of 6 Station / Unit Eshibit E 10CFR50.59 SAFETY EVALUATION 1.
List the documents implementing tbc proposed change.
Miuot PLAWT C H A9GE. Pc4 -t-%l27 B -1 d7C i B - u-11. B -l469,4 B - l4G7 DLr G 2.
Describe the proposed change and tbc reason for the change. A UD /c> c.
BolTFP THREAPED ROD luSTALL NE W RE fL ACEME uf WELM P LocA1cp To RE.fLACE-2005 AM CHoct WCy THe T oueucacra. 5 u ppocTs fM THE Tocus.
A 5 A(hplE. OF RODS 16 BE NG RE.M OV EP Foc. E x A tm 9 ATiou To C ow f lRm THe Asseucs OF S tress coccoslow CRACKIMS F
3.
Is the change:
N Permanent
[ } Temporary -
Expected duration AND Plant Mcde(s) restrictions widle lautabd (NONE if no plant mode restrictions apply)
List tie SAR sedions wideh describe tbc affeded systems, structures, or compoecats (SSCs) or as list the SAR accident analysis seaions which Av=== the affeded SSCs or their operation. Ust any 4.
controtting documents such as SERs, predous modifications or Safety Evalnaw=ma, etc.
Dis c H A RC,G - LIFSAE SECT. 442 h1Alu STe Arn L 19 e es LicF VALVE C H Amsnc boeus),uFSAR SECT. 522.
PRESSURE SufEESSlow e
.y OE41(40)
r Mod a Po4 9i-127 Edibit E ENC OE-06.1
'~
Revision 3 Station / Unit O t) A D ClTIE S [I Page 2 of 6 Exhibit E 10CFR50.59 SAFETY EVALUATION 5.
Describe how the change will affect plant operation when the changed SSCs function as intesded (i.e., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes.
loclude a discussion of any changed interactions with other SSCs No C H A u s e.
To opegglioy DISP Ac6p A S A RE50L7 L
THe-ve ey SMA LL A mount OF W AT E R.
oF T Hts MPc 16 NOT osse esAs.LE. WITHlu THe A C C U RA CY OF Ev.tSTING WAT6 c. LE. VEL N mea 7oc5, 6.
Describe bow the change will affect equipment failures. In particular, describe any new failure modes and their impact during all applicable operating modes.
THe es is m
OPGf2ATtWG E~cuifMrMT INvolveo le THis Wogg.
7.
Identify each accident or anticipated transient (i.e, large/small break LOCA, loss of load, turbine missiles, fire -
flooding) descibed is the SAR where any of the following is true:
The change alters the initial conditiona used in the SAR analysis d
ne changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or failure of the changed SSC could lead to the accident ACCIDENT SAR SECTION Nokie.
8.
Ust each Technical Specification (Safety limit, Umiting Safety System Setting or limiting CWh for Operation) where the requirement, associated action items, associated surv- -
, or bases may be affected.
klo9e THe TEcHmcAl spe e ific Aliou MAgnom AND hwom LJATsct Levels IM The pct ssuct S vppcE ssiop C H A MBE l2 A p.e mot AFFEC-Teo BY THE VERY S M A LL AMOV4T OF W AT s c-Dis ptAcep B y T His N fe.
j f
-nQ.
OE-06.l(41) 1
uoJ o Po 4-1-9H3 Exhibit E ENC OE 06.!
Revision 3 Station / Unit @U AD CITIES /l Page 3 of 6 Erbibit E 10CFR50.59 S AFETY EVALUATION 9.
Will the change involve a Technical Specification revision?
[ ] Yes NNo if a Techolcal Specification revisico is involved, the change esopot be implemented until the NRC issues a license amendment. %)en compledag Step 14,ledicate that a Teebelcal Specifiestjon revistos is required.
10.
To determine if the probability or the consequences of an neddent or malfunaion of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answr the following quesdons for each accident listed in Step 7. Provide the radonale for all NO answers.
Affeded aeddent MO MG-SAR Section:
May the probability of the accident be increased?
[ ] Yes hNo Tac mionum Re suicc o rec 3on.
oF SAFETY Fet Tu Couuncatou H AS NOT BEE M CH ANG ED 50l T H THE VIE
[
oF AwY OF THE Me ta REPLAccmeMT Asse:meurs.
May the consequences of the acddent (off site dose)
[ ] Yes h No be increasedt e
L A NIA Y515.
THIS C HAuce HAS No EFFECT ON A CCi PEUT May the probability of a malfunction of equipmcat
[ j Yes No important to safety increase?
Tr e miuimum ci.s uir2 c D t ACToe OF CATETy feC THE c ou v E c-1100 9A5 Ocr EtFN t HAuctn p il H THE USE OF Auy or 1 at NN Mfl Ace-Me N1 AssEMBUES May the consequences of a malfe*= of equipment
[ ] Yes h No
' mportant to safety locrease?
a Tms epuGt-HAs No tyrecT ev S AFe Ty gg L AT E D E&ulf me uT ofeCNitou if any an to Onesdoe 10 is YES. then an Unmiewed Safety Questies exists.
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Exhibit E 10CTR50.59 SAFETY EVALUATION 11.
Based on your answers to Questions 5 and 6, does the change adwrsely impaa systems or funaions so as create the possibility of an acddent or malfunction of a type difTerent from those evaluated in the SAR7
{ j Yes No Describe the rationale for your answer.
4MAWNG DOES NOT AFF667 Ed?u/PMB/T OPE 467/cNS OR FtfMficNS M,
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Page 5 of 6 1
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i Exhibit E 10CFR50.59 SAFETY EVALUATION 12.
Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer abe following questions for each Technical Specification listed la Step 8. If no Technical Specifications are impacted, then no reduction in neargin of safety czists - proceed to Step 14.
Techaie=1 Specification MN/
Determine which of the fo8owing is tree for the above aparde=eia=-
AR changes to the paramasers or ea-diela== esed to estabEsk the Technical W requirements are is a commervatne direction. hrdars, the actual =~=r*=== Esmit ased not be id==sifiad to determine that no th in margin of safety esimes - proceed to Question 13.
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h Technical (p,,ir,'== provides a margin of safety or==temaar Emit for ebe appEcable paramusar or = =diela= List the immis(s)/margim(s)~belour.
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' M appEcable paramaser or aa-diesa= change is in a potestinBy nom. conservative dronion and the Technical trarifie=ela= meister provides an==gn=== Emit nor anyEciety refusemens a Esmit in the SAR.
Request Nedear Ucessing==s*==ae to idesafy the==y*- Essit/marps for ths Margin of Safssy daer- - eia= use the Emit (s)/mesks) besoar.
Ust Awa=ar Unit (s)/Margim(s) of Safmy i
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13.
Use the above liasies to determine if the snargin of salmy is redmond (la, the aserinimes emesed the acagsames limits). Descrbe the rationale for your determination. Indade a description of compemesalag funors used to reach that cance==ian NO ZHAWE TO 76*Cy'fiV/dt4L Vfc/FM,41/0ff.6 s
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Aw a Po4-l c3 t-m7 Edibit E ENC-OE-06.1 Redsion 3 Station / Unit bAM O[/6.6[l Page 6 of 6 I v Exhibit E 10CFR5039 SAFE 1Y EVALUATION 14. Check one of the following- [] An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. m proposed champ MUST NOT be implemented without NRC approval. ] No Unrewewed Safety Quessina will result ( Steps 10,11, and 13) AND no Technical trar.r-new l revision will be involved. N change may be imple-eared is accordance with applicable procedures. A Terha' =1 Specification revision is invahed; be ao Unreviewed Saimy Question wR resuk. De [] e proposed change requires a License A=*=d=>d Notdy traria= Reguissory Assurance and Nacisar Lkensing that a Technical trar rens6 revision is required. Mark belour as appEcuble. [] The change is not a plant =aM-=ria= or minor plant change and wiB mot be
- under 10CFR50J9. Upon receipt of the approved Technical tracar-=e6 change frees the NRC, the champ muy be implanested.
II ne chaase is a piam =a"-dia-or mi.or piam ch ge. Mark bei= as appEcabia. - is requ* ed. He choses MUST NOT ' " ^ * [] A rension to as castemsTechnical!r - m be '=m ned unta recapt of as approved Technical *, - ssinion. [] ne change wiB mot consist with any emi% Technical ap as-as and omh ase 3~ Ts.:,.i r, _ -' - ^ = required. In them cases, Nedser Ussashg muy asshesims J be not operation, prior to receipt of MRC syysomel af the Unsens e .= .i a d er ch ashorhasion is yand, the nieck bassir shause be sheitsd. [] Nuclear unemming has authorised *h but act operusion, prior to receipt of NRC approval of she Licamme *===d==d hs 30CFRSR.9 Seisay Evolustion ind'tstes that so Unreviewed $sisy Question wE somsk and pro.idos a-ha iry for + 1y. 1-I 'ty rrev=, ~ Sis swe ~e. Q. !. %15. / Does .,,,r s.gd.'z<As%. ~,g-
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