ML20116B039
| ML20116B039 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/08/1996 |
| From: | Pearce L COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LWP-96-048, LWP-96-48, NUDOCS 9607290073 | |
| Download: ML20116B039 (84) | |
Text
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G>mmonwealth lilison Compan)
Quad Cities Generating Station 22710 2(k>th Asenue North I
Girdova, II. 61212 97 40 Tel.4m>s 4-2211
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LWP-96-048 l
July 08, 1996 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 fourth quarter of 1995, for Quad-Cities Station Units 1 and 2, DPR-29 and DPR-30.
A summary of the i
safety evaluations are being reported in compliance with 10CFR50.59 and 10CFR50.71(e).
Respectfully, Comed Quad-Cities Nuclear Power Station D. B.Jud f L. W.
Pearce Station Manager LWP/dak l
l Enclosure cc:
H. Miller, Regional Administrator C. Miller, Senior Resident Inspector 1
STMGRWO48961%T l
9607290073 960700' PDR ADOCK 05000254 R
PDR A Unicom G>mpan)
Spent Fuel Pool UFSAR Change DESCRIPTION:
This change to the UFSAR incorporated the analysis and provisions as a result of Spent Fuel Pool modification.
A portion of this UFSAR change is to allow the fuel pool to be l
cooled using only the fuel pool cooling heat exchangers at i
an appropriate time with a full core off load in the fuel I
pool.
The modification analysis indicated that in the case of a full core off load, one RHR heat exchanger be available for use with a flow of at least 1000 GPM in conjunction with two fuel pool cooling heat exchangers, if needed.
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:
l The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is l
explicitly or implicitly assumed to function during or j
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 Fuel Pool Cooling UFSAR SECTION:
9.1.3.3 For this accident, it has been determined that the change l
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 l
evaluated in the UFSAR.
l 2.
The possibility for an accident or malfunction of a l
different type than any previously evaluated in the UFSAR is
)
not created because the RHR system will be available to cool the fuel pool with a full core discharge until the point I
where the fuel pool cooling system alone is adequate to i
maintain the pool temperature less than the normal discharge licensing acceptance limit.
The implementation procedures, primarily QFP 100-1, will require an evaluation to verify j
that the pool temperature can be maintained less than the normal discharge licensing acceptance limit using only the fuel pool cooling system prior to removing both RHR systems i
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SE-95-082 CONTD from service.
This change provides for adequate cooling of the fuel pool with a full core discharge consistent with the analysis performed in the Spent Fuel Pool Modification and approved in the NRC SER.
No changes to the systems lineup or design conditions are involved.
The change, therefore, does not create the possibility of an accident or malfunction 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.
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SE-95-059 QCTS 360-2 DESCRIPTION:
l FSAR Section 7.7.4.2 described the Electro-Hydraulic Control l
(EHC) System in particular pressure regulator control.
This section stated the setpoint of the backup pressure regulator l
is normally set at 10 psig above the setpoint of the i
operating pressure regulator.
It was proposed to change the l
pressure regulator setpoint to 3.0 psig which was believed to lessen the affects of a regulator failure.
It was also proposed to change this section of the FSAR to state a l
normal range for the backup pressure regulator between 3.0 and 5.0 psig, with a lower extreme limit of 1.0 psig and an upper extreme limit of 10.0 psig.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine l
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 l
UFSAR analysis.
l 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 accident which meet these criteria are listed below:
Steam Pressure Regulator Malfunction UFSAR SECTION: 15.2.1 For this accident, 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 l
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 this change is in a more conservative direction and will not introduce any new failure types.
Reducing the pressure setpoint to 3.0 psig will cause less control valve demand upon a pressure regulator failure and thus lessen the effects on the reactor (pressure and power spike) if a pressure regulator failure were to occur.
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SE-95-059 CONTD 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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M04-1-93-007A RVWLIS Continuous Backfill System DESCRIPTION:
Modification M04-1-93-007A-1 installed a continuous backfill system for the narrow, medium and wide range of RVWLIS.
This was performed by providing a tie-in into the CRD drive water header for a supply of high pressure, deaerated, clean i
demineralized water.
The tie-in occurs between the drywell i
penetration (X-108 & X-109) and the root isolation valves (1-0263-2-12A(B), 1-0263-2-18A(B) and 1-0263-2-41A(B)).
The backfill water is regulated and monitored for flow at a new panel rack near the 2201-5 and 2201-16 instrument racks.
These new racks will be named the "RVWLIS CONTINUOUS BACKFILL FLOW STATION" and labeled 1-2201-5X and 1-2201-6X.
Each flow station will be located inside the caged area of racks 2201-5 and 2201-6.
At each of the flow stations, there will be a split to 2 flow regulating stations, 1 for each of the two reference legs.
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,:.11P.
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:
i Instrument Line Break UFSAR SECTION: 15.6.2 Loss of Coolant Accident UFSAR SECTION: 15.6.5 Increase in Feedwater Flow UFSAR SECTION: 15.1.2 For each of these accidents, it has been determined that the change described above will not increase the probability of l
an occurrence or the consequence of the accident, or l
malfunction of equipment important to safety as previously evaluated in the UFSAR.
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l M04-1-93-007A CONTD 2.
The possibility for an accident or malfunction of a different type than any previously. evaluated in the UFSAR is not created because a common mode failure of all instrumentation on a single or multiple reference legs is not an accident discussed in the FSAR/UFSAR.
A USQ submittal by licensing requests NRC review of the use of check valves to maintain water level in the reference legs.
It is necessary to be certain that a gross failure of the CRD system will not cause redundant instrument loops to fail due to check valve failures.
This USQ will be addressed with the licensing amendment of the Technical Specification changed.
There was also a U?n submittal (by CECO) for the accidental closure of reference leg root valves with the Backfill Subsystem operating.
This condition created a plant transient due to the Backfill Subsystem pressurizing the isolated reference leg.
This USQ is no longer required, i
because the injection location has been changed by Addendum l
- 2.
Closure of the root valve will no longer pressurize the reference leg.
l The NRC.has approved the modification for temporary use on l
U2 only.
Reference Amendment No. 139 to DPR-30.
A CECO i
commitment was made to change the design to no longer rely on administrative controls to avoid a plant transient.
This addendum satisfies that commitment to the NRC.
Other system failure modes and interfaces have been evaluated to'show that they will have no significant detrimental affect on Safety Related equipment or equipment important to plant safety.
These interfaces and failure modes have been evaluated by the Designer and considered in the sizing of equipment components.
l 3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because additional leak path must be reviewed by the NRC in a Technical Specification change.
The installed configuration must also be evaluated by the NRC in accordance with 10CFR50, Appendix A (GDC 55).
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SE-96-034 UFSAR Change No. 96-50 i
DESCRIPTION:
Revised Section 17.2-1 of UFSAR to delete reference to procedures being documented within the Comed Quality Assurance Program Manual.
Also revised Section 17.2-1 of the UFSAR to delete reference to delineation between Comed stations the applicable Reg Guides and ANSI Standards.
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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, l
or component could lead to the accident.
The accidents which meet these criteria are listed below:
None l
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is l
not created because the change does not affect systems or functions.
The change is an administrative change to reflect the contents of the current =QA Manual.
The change does not create the possibility of an accident or malfunction of a type different from those evaluated in the l
i 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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E04-2-93-323 MOV 2-1402-24 A(B) Code Reconciliation /TOL Testing DESCRIPTION:
Valves MO2-1402-24 A(B) are the inboard injection valves for the Core Spray System.
These motor operated valves normally isolate the Core Spray System from the reactor vessel.
The valve is automatically opened when the Core Spray System is initiated and reactor pressure becomes less than 325 psig.
The valve can be throttled to control system flow below 4500 9pm.
The MO2-1402-24A(B) valves are interlocked in the closed position when reactor pressure is greater than 325 psig.
Additionally, the valves are interlocked with their associated outboard injection valves [MO2-1402-25A(B)] to 1
prevent open.ing both injection valves simultaneously at high j
reactor pressures.
l This Exempt Change increased the thrust output of the valve actuators to ensure the valve can be positioned against j
system design conditions.
This Exempt Change also increased the Structural Thrust Limit of the valves by replacing limiting components with new components having higher thrust limits.
To increase the valve actuators thrust output the overall gear ratio of the actuator was changed from 29.44 to 59.4.
Additionally, the existing 3600 rpm, 60ft-lb actuator motor was replaced with a new 1800 rpm, 80ft-lb motor.
New cabling was also installed to facilitate the new motor loading.
These changes increased the valves stroke time from 9.9 seconds to 40 seconds.
j To increase the valves Structural Thrust Limit the existing disk and stem was replaced with new components with higher structural thrust limits.
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.
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E04-2-93-323 CONTD 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 Coolant Accidents SAR SECTION:
15.6.5 Resulting from Piping Breaks Inside Containment For this accident, 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 SAR.
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 failure modes or system interfaces are created.
These changes will not modify the function of the valves,. nor will they affect any interactions with other safety related components or systems.
Therefore these changes will not create a new accident scenario or malfunction not already 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.
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SE-95-039 UFSAR Change Request DESCRIPTION:
UFSAR table 6.2-7 is being updated to reflect that certain Containment Isolation Valves (CIV's) also serve a Pressure Isolation Valve-(PIV) function.
A note is being added to the table which will designate the 1402-9A/B, 1402-25A/B, j
1001-68A/B, 1001-29A/B, 1001-47, 1001-50, 220-44, and 220-45 valves as PIV's in addition to CIV's.
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 j
UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
l Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
None 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this change only provides an update to UFSAR Table 6.2-7.
No plant equipment is being modified.
No new testing is being developed.
The testing requirements for the PIV's have previously been determined and all of the PIV's are already being tested per the IST program.
No new accidents or malfunctions are being created by this UFSAR update.
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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k M04-2-93-012 Replace Dresser Electromatic Relief Valves with Target Rock Power Operated Relief Valves.
DESCRIPTION:
The purpose of this modification was to replace the Main Steam Relief valves.
The existing Dresser Electromatic relief valves (ERVs) were be replaced with Target Rock power operated relief valves (PORVs).
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:
Increase in Heat Removal SAR SECTION:
15.1 by the Reactor Coolant System Decrease in Heat Removal by the Reactor Coolant System SAR SECTION:
15.2 Decrease in Reactor Coolant Inventory SAR SECTION:
15.6 Anticipated Transients j
15.8 j
i For 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 SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is J
not created because the valve replacements will not 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.
The replacement Target Rock PORVs serve the same function as the existing ERVs.
The new valves are designed, constructed, and qualified to ASME Section III, and IEEE requirements for relief valves operating in harsh environments.
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M04-2-93-012 CONTD l
They are also qualified for normal operating loading conditions (dead weight and thermal), seismic conditions (OBE & SSE) and dynamic conditions from safety or relief valve opening.
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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E04-1-93-216 Core Spray Testable Check Valves DESCRIPTION:
Modified Core Spray Testable Check Valves AO 1-1402-9A/B to remove the test functions due to numerous problems with pneumatic actuators, remote position indicators and packing leaks.
The pneumatic actuator limit switches, local push-buttons & associated wiring were removed.
Each valve body had a conversion kit installed due to the removal of the test function.
Other electrical demolition work included de-termination and removal of wiring internal to panel 910-32 & 901-33, and sparing relays 1430-119A/B.
At control y
panel 901-3, the check valve push-buttons, disk position l
indicating and actuator position lights and associated internal wiring were removed.
Other associated i
conductors / cables were spared at panels 901-3, 901-21, 901-32, 901-33 and penetrations X-100B, X-100G And x-103.
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SAFETY EVALUATION
SUMMARY
1 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:
)
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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:
15.6 l
For this accident, it has been determined that the change described above will not increase the probability of an i
occurrence or the consequence of the accident, or i
malfunction of equipment important to safety as previously evaluated in the SAR.
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I E04-1-93-216 CONTD l
2.
The possibility for an accident or malfunction of a l
different type than any previously evaluated in the UFSAR is not created because the modified check valves will perform the same function as before, therefore this change is within the boundaries of the UFSAR.
As was previously stated, the l
removal of the air operator has no adverse functional impact l
on the operation of the check valves.
The change will not I
create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR.
l 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not l
reduced.
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M04-2-92-019 I
DESCRIPTION:
Quad Cities has experienced several Group I isolations following turbine trips.
These isolations have been l
attributed to spurious Main Steam Line (MSL) low pressure (LP) signals which occurred while MSL pressure appeared to l
remain above the 850 psig setpoint.
The Technical Specification setpoint for this instrument is 825 psig.
l Special Test 2-104, performed 1/1/92, has confirmed that these spurious isolation signals are caused by short l
duration pressure oscillations in the sensing line which are l
initiated by pressure transients in the MSL upon closure of l
the Main Turbine stop Valves.
The purpose of the MSL LP isolation function is to provide automatic protection against excessively rapid reactor de-pressurization and vessel cool down that could occur during l
the limiting de-pressurization FSAR event, the failure of the pressure regulators in the open direction.
A General Electric analysis of this event has determined that up to a one second delay is allowable without a reduction in the margin of safety.
A 0.5 second delay was conservatively selected so as to resolve the described problem while remaining well within the established one second criteria.
This modification installed an adjustable time delay relay l
in each of four channels of the MSL LP Group I isolation circuitry in order to increase the delay time between sensed l
low pressure and Group I activation, and therefore, l
effectively filter out the observed high frequency pressure l
oscillations.
The relay setpoint has been determined by subtracting the response time of the sensing lines, the pressure switch, the existing circuitry, and the operating errors of the neu I
relay from the 0.5 second acceptance value.
The final relay setting of 0.30 seconds will result in an actual delay of 320 msec (+21.18/-11.18 msec).
The associated reset time of 0.025 seconds is short enough to adequately filter out the MSL sensing line oscillations.
In the case of a pressure oscillation, the delay cycle of the new relay will be reset before the delay setpoint has been reached.
However, a MSL LP PCI will be initiated if sensed pressure in the main steam lines is below the low i
pressure setpoint for longer than the time delay.
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M04-2-92-019 CONTD l
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SAFETY EVALUATION GUMMARY:
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the i
l UFSAR where any of the following is true:
The change alters the initial conditions used in the l
UFSAR analysis.
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The changed structure, system or component is l
explicitly or implicitly assumed to function during or after the accident.
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Operation or failure of the changed structure, system, or component could lead to the accident.
1 i
The accidents which meet these criteria are listed below:
Steam Pressure Regulator Malfunction SAR SECTION:
15.2.1 l
For this accident, it has been determined that the change j
described above will not increase the probability of an l
occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the SAR.
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 new relays will be identical in operation to the existing relays, with the exception of the time delay that they will introduce.
However, this time i
delay does not change the design basis or create any operating conditions that are outside the previously l
analyzed set of safe limits.
l 3.
The margin of safety, as defined in the basis for any l
Technical Specification, is not reduced because the low pressure setpoint will not be changed.
The modification introduces time delay in the actuation logic.
When added to the existing logic delay, the total logic delay time will be 500 msec or less.
This delay time is well within the one second delay assumed in the original design basis.
Therefore, the new value will not exceed the acceptance limits and the margin of safety is not reduced.
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DESCRIPTION:
l Quad Cities has experienced several Group I isolations following turbine trips.
These isolations have been attributed to spurious Main Steam Line (MSL) low pressure (LP) signals which occurred while MSL pressure appeared to remain above the 850 psig setpoint.
Special Test 2-104, performed 1/1/92, has confirmed that these spurious isolation signals are caused by short duration pressure oscillations in the sensing line which are initiated by pressure transients in the MSL upon closure of the Main Turbine Stop Valves.
l The purpose of the MSL LP isolation function is to provide l
automatic protection against excessively rapid reactor de-l pressurization and vessel cool down that could occur during the limiting de-pressurization FSAR event, the failure of the pressure regulators in the open direction.
A General Electric analysis of this event has determined that up to a l
one second delay is allowable without a reduction in the margin of safety.
A 0.5 second delay was conservatively selected so as to resolve the described problem while remaining well within the established one second criteria.
l This modification installed an adjustable time delay relay in each of four channels of the MSL LP Group I isolation circuitry in order to increase the delay time between sensed low pressure and Group I activation, and therefore, effectively filter out the observed high frequency pressure oscillations.
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.
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Operation or failure of the changed structure, system, l
or component could lead to the accident.
l The accidents which meet these criteria are listed below:
f Steam Pressure Regulator Malfunction SAR SECTION:
15.2.1 Q:\\1TCHOP3\\SAfflT\\l96RPT.wpf
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N04-1-92-019 CONTD l
For this accident, it has been determined that the change l
described above will not increase the probability of an l
occurrence or the consequence of the accident, or l
malfunction of equipment important to safety as previously j
evaluated in the SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the new relays will be identical in design and performance to the existing relays, with the exception of the 300 maec time delay that they will introduce.
However, this time delay does not change the design basis or create any operating conditions that are outside the previously analyzed set of safe limits.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the low pressure setpoint will not be changed.
The modification introduces a 300 msec delay in the actuation logic.
When
.added to the existing logic delay, the total logic delay time will be 500 msec or less.
This delay time is well within the i sec delay assumed in the original design basis.
Therefore, the new value will not exceed the acceptance I
limits and the margin of safety'is not reduced.
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SE-96-026 Use of simplified one Line Drawings in UFSAR Versus P& ids DESCRIPTION:
The proposed activity is a revision to the UFSAR, substituting simplified, single sheet flow diagrams for existing Piping & Instrumentation Drawings (P& ids).
Table 1 lists the UFSAR figures that are being removed and identifies the number of their corresponding replacement figures (various P& ids are not being replaced which is also noted).
Reduction of UFSAR figures will reduce the administrative burden of submitting the latest revision of one-hundred seventy-six (176) P& ids bi-annually.
This activity produces a simplified drawing where an entire system is presented on a single drawing.
This was accomplished for twenty-four systems.
The existing UFSAR P& ids are based on the M P&ID drawings.
M drawings are somewhat elaborate piping schematic drawings that show such detail.as line sizes and classifications.
The new, simplified drawings were reviewed by System Engineers.
Regulatory Guide 1.70, Rev.
3, Nov. 1978, was reviewed to establish which drawings should be maintained in the UFSAR.
Reg. Guide 1.70 states, " Drawings, maps, diagrams, sketches, and charts should be employed where the information can be presented more adequately or conveniently by such means."
8AFETY 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:
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r SE-96-026 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 activity is not a physical modification to the plant.
Likewise, there is no change l
created between the physical plant, and the SAR description of the design, function, or method of performing the function of various systems, structures, and components (SSC).
This safety evaluation concerns the removal of i
Piping and Instrumentation Diagrams (P& ids) used in the i
UFSAR to provide the user a visual reference that compliments the textual description of the various SSCs in i
the UFSAR.
All changes, tests, and experiments will continue to be I
I evaluated in accordance with 10CFR50.59.
The information being removed from the drawings'may be implicit in the UFSAR text description of various SSCs.
The guidance used to l
determine which drawings to include as identified in NRC l
Regulatory Guide 1.70, Revision 3, and CECO (including Licensing) requirements remains.
When future changes are made associated with this information, the effect on the SAR description of the design, function, and method of performing the function of the SSCs explicitly described will be evaluated to determine if a UFSAR change and associated Safety Evaluation is necessary.
If a component is a part of a parent system, and modifying that component results in the. parent system operating or functioning contrary to the parent system's SAR description, then a Safety' Evaluation will be performed.
Therefore, the possibility of a malfunction or accident of a different type than any previously. evaluated in the SAR is not created.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because this activity is not a physical modification to the plant.
Likewise, there is no change created between the physical plant, and the SAR description of the design, function, or l
method of performing the function of various systems, j
structures, and components (SSC).
This safety evaluation concerns the removal of Piping and Instrumentation Diagrams i
(P& ids). used in the UFSAR to provide the user a visual reference that compliments the textual description of the various SSCs in the UFSAR.
i QNniCHOPS\\$APETYut6kFT.wyf.
l l
l l
SE-96-026 CONTD l
All changes, tests, and experiments (CTEs) will continue to be evaluated in accordance with 10CFR50.59.
The information being removed from drawings may be implicit in UFSAR text description of various SSCs.
The guidance used to determine which drawings to include as identified in NRC Regulatory Guide 1.70, Revision 3, and CECO (including Licensing) requirements remains.
When future changes are made associated with this information, the effect on SAR description of design, function, and method of performing l
the function of the SSCs explicitly described will be 1
evaluated to determine if a UFSAR chant,b and associated l
Safety Evaluation is necessary.
If a Eomponent is part of a l
j parent system, and modifying that component results in the l
parent system operating or functioning contrary to the parent system's SAR description, then a Safety Evaluation will be performed.
The Margin of Safety defined in Tech Specs is unchanged.
Therefore, the margin of safety as defined in the basis for any Technical Specifications is not reduced.
i l
I l
l l
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l M04-2-95-006 Scram Discharge Volume (SDV) Logic Design Change DESCRIPTION:
A modification is required to restore the Scram Discharge Volume (SDV) logic to the requirements outlined in IEEE 279-68 single failure criteria.
Two new junction boxes were mounted near the existing junction boxes on both the north and south Scram Discharge Volumes (SDV).
The RIS level l
switches were removed from the existing boxes and mounted in l
the new junction boxes.
Conduit and cables were reworked as required to support the relocation.
For the north side the new boxes supply sub-channels Al and Bl.
The south side junction boxes provide trip inputs to subchannels A2 and B2.
Computer inputs from the north SDV were disconnected and l
abandoned to provide proper divisional separation.
These computer points are non-safety related and used for historical trending only.
They do not provide any input or function other than indication to the process computer.
l Future reconnection of these computer points may be considered as a separate design change but will not be l
addressed by this modification.
l 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.
The accidents which meet these criteria are listed below:
1 ATWS SAR Section 15.8 For this accident, 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 SAR.
I 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is i
not created because the change will allow the equipment to function as it was intended per the UFSAR.
Q:\\1TfHOP3\\SAIUYu96RIT.wpf
M04-2-95-006 CONTD 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
i
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l QanCHOP5\\SAFEmlMRFT.wpf
I i
UFSAR TABLES 8.3-1 AND 8.3-3 UFSAR Change for EDG Loading DESCRIPTION:
The peak ratings for the EDGs given in the UFSAR are given as 2860 KW and 3575 KVA.
The actual value as specified by the vendor is 2850 KW and 3560 KVA.
Table 8.3-1 needs to be changed to list the peak rating as 2850 KW and 3560 KVA.
Table 8.3-3 needs to be changed to list the peak EDG load as 2850 KW.
8AFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the i
l UFSAR where any of the following is true:
l I
The change alters the initial conditions used in the UFSAR analysis.
The changed structure, system or component is l
explicitly or implicitly assumed to function during or l
l after the accident.
Operation or failure of the changed structure, system, l
or component could lead to the accident.
The accidents which meet these criteria are listed below:
SAR Section 8.3 l
For this accident, it has been determined that the change l
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 SAR.
2.
The possibility for an accident or malfunction of a
[
different type than any previously evaluated in the SAR is l
not created because this change restores the peak load on the EDGs to its original design basis of 2850 KW.
The i
accidents evaluated in the SAR were originally evaluated considering a peak EDG loading of 2850 KW.
Since this change restores the peak loading to its original design basis, there are no new possible accidents created.
3.
The margin of safety, is not defined in the basis for any Technical Specification, thet c(cre, the safety margin is not reduced.
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, _ __._ _ _ _ m.__
l t
E04-1-93-327 i
DESCRIPTION:
E04-1-93-327 reinstalled the transformer control circuitry and evaluate the electrical compatibility of the new transformer.
These changes are necessary due to slight differences between the GE and SMIT transformers.
The control circuitry changes will not affect the operation of the plant.
SAFETY EVALUATION
SUMMARY
1 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.
I 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 SAR Section 8.3.1 For this accident, it has been determined that the change i
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 SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the SAR 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 came 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 presiously evaluated in the 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.
carscuomSAFEnfuMRMspf
E04-2-93-198 Feedwater Regulating Valve Actuator Replacement DESCRIPTION:
E04-2-93-198 installed new hydraulic actuators on feedwater regulating valves (FRV) 2-0642A/B.
A new fire protection system was installed to mitigate the addition of a large amount of combustible material (80 gallons of hydraulic oil).
A new drainage and curbing barrier system will remove water during a suppression system actuation and remove oil should a pipe rupture.
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.
i Operation or failure of che changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Increase in Feedwater Flow SAR Section 15.1.2 Loss of Feedwater SAR Section 15.2.7 For these accidents, it has been determined that the change described above will not increase the probability of an l
occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the SAR.
1 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the SAR is not created because the operation of the new valve operators will be~ identical to the existing operators.
The new operators are more reliable and accurate.
The only changes that could adversely affect other equipment is the addition of 80 gallons of hydraulic oil (fire hazard).
The addition of the fire hazard has been mitigated by installing a fire detection / suppression system.
The new valve operators are designed to lock in place upon the detection of any operational errors, which is consistent with the existing 1
operators.
No new accidents or malfunctions have been identified that have not been evaluated.
l I
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E04-2-93-198 CONTD 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
I l
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SE-96-011 QCEPM 700-03 l
l DESCRIPTION:
Surveillance QCEPM 700-3 was revised to discontinue the present inspection requirements, except for the coil l
inspections of normally energized relays with lexan or nylon bobbins.
A inspection was added to QCEPM 700-3 to sample the relay l
contact resistance.
A sample size of 40 relays per unit i
will be inspected.
This resistance check will be done on one contact of each of the 40 relays.
Acceptance criteria i
for this check is 1 Ohm (based on current station procedures).
If the contact resistance check indicates l
oxidation, the sample size will be increased.
l UFSAR section 7.2 paragraph 55 will be revised, from:
UFSAR section 7.2 paragraph 55 states " Reactor protection system safety-related HFA relay inspections are presently conducted during each refueling outage.
These inspections confirm cleanliness of the relay pole components and verify that the coils are not deteriorating."
To:
UFSAR section 7.2 paragraph 55 will state " Reactor protection system safety-related HFA relay inspections are not required.
These relays have had their coils replaced with General Electric century series coils.
This replacement ensures continued reliable operation of these relays."
Note:
The suspect RPS relay coils were noted as having a lexan or nylon bobbin by G.E.
The suspect relay coils have been replaced with G.E.
century series coils.
The new coils have a tefzel bobbin.
These coils corrected the deterioration problem.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine i
each accident or anticipated transient described in the l
UFSAR where any of the following is true:
The change alters the initial conditions used in the UFSAR analysis.
4 The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
i Q:miCHOP3\\SAMT. Y\\l96RJT.wpf 4
l SE-96-011 CONTD l
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
l ATWS UFSAR Section 7.2, 4.6, 7.7, 15.8, 7.6, 7.8, 6.2, 5.4, 9.3, and 7.0 LOCA UFSAR Section 9.3, 6.3, 6.2, 5.4, 8.3, 7.3, 5.2, 3.0, l
5.0, 3.9, 3.11, 15.3, 3.8, 15.6, 9.5, 6.4, 4.6, and 4.4 l
Main Steam Line Break
(
UFSAR Section 15.6, 5.4, 10.3, 7.6, 6.2, 15.0, 5.0, l
3.9, and 3.6 High Energy Line Break l
UFSAR Section 3.6, 3.8, and 6.3 For 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 the change in the procedure will not l
adversely impact systems or system functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR.
The change only affects the inspection of the relay, not the function of the relay.
The revised inspection incorporates a contact oxidation check using MIL-STD 105E for development of the sample size.
This MIL-STD is an NRC accepted method for development of a sampling plan.
This inspection will ensure that relay failure due to contact oxidation will be predicted prior to l
failure with a 95% confidence level.
l l
The removal of the visual inspection for the total population is due to coil replacements.
The reason for inspecting these relays was that the coils were deteriorating in a manner that would prevent the relay from performing its intended design function.
This deterioration is described in General Electric SILs S332, S332S1, SO44, SO44S1, SO44S2, SO44S3, SO44S4R2, SO44SS, and S153.
When the suspect relay coil is replaced with a G.
E. century series coil, the identified failure mode is removed.
The replacement coils were supplied in accordance with 10CFR21.
This requires the vendor (G.E.) to notify the customer Q:\\TECHOP3\\5 ATE!Y\\t96RPT.wpf
SE-96-011 CONTD (Commonwealth Edison) of any known product defects.
This ensures that the relay coils will not prevent the relays from performing their intended design function.
The relays that have not had their coil replaced are listed in ER9500358.
These relays still require the visual inspection each refuel outage.
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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l SE-95-068 UFSAR Section 7.5-15 (7.5.3) l DESCRIPTION:
6 Delete text from the UFSAR, pages 7.5-8 and 7.5-9, that explains the operation of the Safety Parameter Display System (SPDS).
8AFETY EVALUATION
SUMMARY
i 1.
The change described above has been analyzed to determine each accident or anticipated transient described in the l
UFSAR where any of the following is true:
l l
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 l
after the accident.
l Operation or failure of the changed structure, system, or component could lead to the accident.
i The accidents which meet these criteria are listed below:
None 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because by deleting the text in the UFSAR that explains the operation of SPDS does not adversely impact systems or functions so as to create the possibility of a malfunction of a different type from those previously evaluated in the UFSAR.
Operation of the SPDS does not change.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
I 1
l QdTFrHOPh5AFITY\\t96RIT.wpf
SE-96-015 Supplement to SE-94-066 per NTS Item #254-350-95-03800 DESCRIPTION:
The changes will affect-the procedure controls for the stability exclusion regions.
Specifically, the immediate exit region will be changed from the region bounded by-greater than 80% Flow Control Line(FCL) and less than 45%
Total Core Flow to greater than 70% FCL and less than 45%
total core flow, there will be a new immediate scram region bounded by the greater than the 100% FCL and less than 40%
total core flow, and there will be an immediate manual scram for operation in the RUN mode on natural circulation.
8AFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine j
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:
Decrease in Feedwater Temp UFSAR Section 15.1.1 Loss of Condenser Vacuum UFSAR Section 15.2.5 Loss of Normal Feedwater Flow UFSAR Section 15.2.7 Loss of Single and Multiple Recirc Pumps UFSAR Section 15.3.1 Recirculation Flow Controller Failure UFSAR Section 15.3.2 Recirculation Pump Shaft Seizure UFSAR Section 15.3.3 Recirculation Pump Shaft Break UFSAR Section 15.3.4 HPCI Initiation During Operation UFSAR Section 15.5.1 For 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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l SE-96-015 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 these procedure changes strengthen the provisions and controls in place to avoid core instabilities.
These changes do not introduce any new accident or transient of a type different from those r
i i
described in the UFSAR.
The changes provide more i
conservative operating instructions should certain l
transients or accidents that are conducive to instabilities l
occur.
3.
The margin of safety, as defined in the basis for any l
Technical Specification,.is not reduced because the margin of safety is increased with these procedure changes and administrative controls.
Thermal hydrolic instabilities have the potential of violating the MCPR safety limit.
These changes decrease the possibility of an instability l
occurring that could challenge the safety limit.
i I
l l
1 I
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l SE-96-005 QCOP 1900-22 " Fuel Pool Cooling Pump Discharge X-Tie" l
DESCRIPTION:
A procedure has been created to cross-tie both Fuel Pools in order to-use the opposite unit Fuel Pool pumps and/or RHR assist for the unit that is performing a core offload.
QCOP 1900-22, Fuel Pool Cooling Pump Discharge Cross-Tie, is regarded as an additional contingency measure to add to operators options and flexibility in the event that a Fuel Pool cooling system failure occurs.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the l
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.
l The accidents which meet these criteria are listed below:
Loss of Fuel Pool Cooling UFSAR SECTION: 9.1.3.3 For this accident, 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_the addition of this procedure will not affect the performance of any other system in the plant.
Other than the cross-tie interconnection of the fuel pool systems, no other electrical or mechanical changes are being implemented.
This procedure will be implemented to maintcin fuel pool temperatures to within-limits described in the FSAR.
f i
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, l
l j
SE-96-005 CONTD The flow capability of the new flow path introduced by QCOP 1900-22 is adequate because it utilizes a 6.0" header which is the size of the rest of the fuel pool system suction piping.
The size of the discharge piping (opposite unit) also remains unchanged, with the only difference being the source of the fuel pool water.
Also, there are no new fuel pool chemistry issues associated with the transfer of fuel pool water as executed in this procedure.
If the' procedure is implemented without the operation of a fuel pool filter demineralizer it would be no different than the current possibility of short term operation without clean-up to achieve proper cooling.
The return flowpath (through the cross-tie canal) maintains the pool inventory for both pools.
The cross-tie canal is normally open between the pools.
~
3.
The margin of cafety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
4 1
l l
QATrCHOP33APEmlMWr.wyf l
SE-95-083 Placing One Hotwell Makeup Pump in Pull-To-Lock DESCRIPTION:
l Revised the UFSAR to place one condensate makeup pump in pull-to-lock, preventing both pumps from auto starting simultaneously on a low condenser water level.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the l
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 l
after the accident.
1
- peration 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 Condenser Vacuum UFSAR SECTION: 15.2.5 & 15.8.5 For this accident, 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 the condensate make-up pumps are I
required to provide make-up water to the condenser when the I
condenser does not have a vacuum or when an increased amount j
of water is needed to maintain hotwell level.
The l
condensate make up pumps do not provide any safety function i
during any accident in the UFSAR.
The loss of makeup capability is the worst case consequence of a failure of the
)
hotwell makeup-pumps.
A loss of makeup capability is l
bounded by a loss of feedwater and therefore does not create the possibility of a malfunction not previously evaluated.
l 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
l l
Q:\\TECHOP3\\SAETY\\l96RPT.wpf
l SE-95-079 UFSAR Changes to Sections 9.5.2.2.1 and Section
13.3 DESCRIPTION
l l
Removed Specific details of the Generating Station Emergency i
Plan (GSEP) from the UFSAR Sections 13.3, 9.5.2.2.1, inserts l
reference statements to the GSEP for specific details.
SAFETY EVALUATION
SUMMARY
t 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 I
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:
None l
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this change does not adversely impact systems or functions since no equipment, structures, l
components or systems are being changed.
The full scope of this change is to remove Specific details of the Generating Stations Emergency Plan (GSEP) from the UFSAR Section 13.3 l
and 9.5.2.2.1 and inserts reference statements to the GSEP for the specific details of the plan.
j 3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
)
i i
1 0:\\TirHOP3\\5AITm196RPT.wpf
SE-95-076 UFSAR Update - Section 7.6.2.3 DESCRIPTION:
This UFSAR change clarifies the purpose of the reactor water level (high) trips.
This design basis statement provides a clearer basis for setpoint calculations.
DAFETY 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.
l l
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.
The accidents which meet these criteria are listed below:
Increase in Feedwater Flow UFSAR SECTION: 15.1.2 For this accident, it has been determined that the change l
described above will not increase the probability of an l
occurrence or the consequence of the accident, or l
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 addition to the UFSAR is clarifying in nature.
There is no change to systems, equipment, or functions.
It merely provides a concise basis for the high level trips.
3.
The margin of safety, as defined in the basis for any l
Technical Specification, is not reduced because the changes to the UFSAR are editorial and do not constitute any change to plant practice or design.
The change gives the purpose of the instruments to that a trip setting can be better justified.
i l
4 i
l Q:\\1TCHOP3\\ SAFE!Y\\l96RPT.wpf
l l
SE-93-131 Exempt Change E04-1(2)-93-118 DESCRIPTION:
Exempt change replaced the existing Molytek DW temperature recorder with a Johnonson Yokogawa.
In addition, the DW temperature indicator and 40 pt toggle switch on the DW Environs Rack was removed.
Rewording of UFSAR Section 5.2.5.2 regarding drywell temperature sensors and recorders will need to be done to support this design change.
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:
None 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this UFSAR change will reflect the :n-built configuration of the plant.
The temperature rec w.ser being replaced on the DW Environs rack is not assumed to function after an accident.
The existing 32 point and new 30 point recorder are to be used for indication only and do not affect any system.
This change will not affect any systems such as to create the possibility of any accident different from those evaluated.
3.
The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
i i
l l
QmOIOmSAITm196RFT.wpf
SE-96-016 Setpoint Change 96-001E & 96-002E DESCRIPTION:
The settings for Time Delay Relays (TDR) 1(2)-0287-120B(C) and 1(2)-0287-121B(C) were increased from 9-10 seconds to 14.5 seconds (1 5).
These relays are designed to prevent the automatic reopening of the ADS valve for a minimum of 10 seconds following an actuation.
j sArary svALuATIOu sexuARY:
l 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 l
l after the accident.
1 Operation or failure of the changed structure, system, i
or component could lead to the accident.
l The accidents which meet these criteria are listed below:
Load Rejection UFSAR SECTION: 15.2.2
]
Turbine Trip UFSAR SECTION: 15.2.3 Inadvertent MSIV Closure UFSAR SECTION: 15.2.4 Loss of Condenser Vacuum UFSAR SECTION: 15.2.5 l
For these accidents it nas been determined that the change described above will not increase the probability of an i
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 the change to the TDRs setpoint will ensure that the ADS logic system will perform as currently described in the UFSAR.
Section 5.2.2.4 describes the function of the ADS logic related to inhibiting-relief valve reactuation.
The UFSAR states "These two valves are l
equipped with additional logic which' inhibit valve reopening (via reactor repressurization or the ADS) for at least 10 seconds following each closure.
(This compares with a F
calculated worst case elevated water leg duration time of 6.3 seconds).
The previous setpoint was adequate 0:miCHOMSAMEIMRM.wyf
\\
SE-96-016 CONTD when compared to the worst case water leg duration time, but did not provide the required design margin indicated 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.
l i
f i
l Q:YlTCHOP3\\SAITIY\\l96RPT,wpf
1 E04-1-93-236 Refueling Platform Upgrade i
DESCRIPTION:
This Exempt Change upgraded equipment and controls on the refueling platform.
The_ work scope included replacement of the power centers, load weighing systems, grapple head, i
bridge position system, auxiliary hoist hose reel and main hoist electrical cable reel.
A detailed description of the specific changes are as follows:
1.
The power center enclosure was replaced with one of similar size and dimension using existing mountings.
a.
A user programmable boundary zone protection system (PBZP) in the power center replaced the existing obsolete zone computer.
The new PBZP system has a bypass push button to allow operations in the boundary regions.
The PBZP system also has provisions to slow the main hoist during hoisting and lowering when mast sections i
" drop-off" and " pick-up" to reduce shock to the platform.
A GE/FANUC PBZP controller replaced the present controller and will provido safety interlocks, fault lock-outs and boundary zone protection.
b.
The previous mast position indication system was a Valuetrol Unit which had to be shared between both bridges since only one Valuetrol was functional.
This was replaced with an absolute position resolver and operator display or each machine axis.
Both the resolver readout and display are dimensionally equal to the units they replace.
c.
The motors and drives for the main hoist, trolley and bridge were replaced.
These motors are a DC regenerative type.
They will be interchangeable and will provide redundant overload protection.
All existing operator controllers will remain the same.
2.
The present load weighing system was replaced with a new solid state system.
The new system uses an electronic compression gage to supply a variable output to the operator's hoist load display, control pendant display, and to the electronics in the power control l
center.
Power center controls will provide the interlocks for adjustable hoist load trip setpoints.
The hoist interlocks are electronically identical to the present interlocks.
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l E04-1-93-236 CONTD l
1 3.
The new grapple head'has a single actuator cylinder and l
sealed magnetic switches for grapple position l
indication open or closed.
A new, more reliable actuator switch and indications for grapple open /
closed was installed.
The new grapple maintains the same " double J-hook" design as the present grapple.
4.
An automatic positioning system was provided.
This system will consist of a touch screen mounted in the operators cab that will allow the operator to select a desired location in the core or spent fuel pool and to command the system to proceed.
The control system moves the bridge and trolley to the selected location.
All pre-programmed boundary zones, interlocks, and speed limitations will be observed.
5.
The auxiliary hoist (frame and mono-rail) had the reel spring packs replaced.
This change reduces the excessive force presently exerted on the spring reels i
and increases the reliability of the air hoses and fittings.
6.
The main hoist electrical cable reel was replaced to reduce the force related failures of the bulkhead connector used on the mast.
There are several reasons for the refuel bridge upgrade.
The bridge has affected critical path in refueling outages because of numerous equipment problems.
These equipment i
problems are due in large part to the fact that the bridge components are over 20 years old.
This bridge upgrade will replace the main sources of unreliable controllers and hardware, thereby effectively eliminating any critical path breakdown.
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 uted in the UFSAR analysis.
l The changed structure, system or component is explicitly or implicitly assumed to function during or l
after the accident.
Operation or failure of the changed structure, system, or component could lead to the accident.
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E04-1-93-236 CONTD The accidents which meet these criteria are listed below:
Mislocated Fuel Assembly SAR SECTION:
15.4.7 DB Fuel Handling Accident SAR SECTION:
15.7.2 (Bundle Drop)
For 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 SAR.
2.
The possibility for an accident or malfunction of a i
different type than any previously evaluated in the UFSAR is not created because existing plant refuel bridge equipment will be replaced with new equipment.
The new equipment i
being installed is more reliable and accurate than the I
present equipment but provides the same function.
Therefore, the design changes do not introduce any new accident or malfunction.
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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SE-96-021 QCTS 340-1 Standby Liquid Control System Outage Surveillance DESCRIPTION:
QCTS 340-1 Standby Liquid control System Outage Surveillance i
has been rewritten to test the system to the new requirements of the Technical Specifications Upgrade Program Y
L (TSUP).
Under the new Technical Specifications,.one of the j
SBLC subsystems, including an explosive valve, will be l
initiated to verify that a flow path from the pumps to the l
reactor pressure vessel is available by pumping l
domineralized water into the reactor vessel.
t The requirement for setpoint testing of the relief valves was also removed from this procedure.
A prerequisite to verify the completion of this testing was added.
Finally the differential pressure limit of 160 paid maximum between the SBLC pumps and the reactor vessel during the 1
l injection test was removed from the acceptance criteria.
]
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine l
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 l
UFSAR analysis.
The changed st-
.ture, system or component is explicitly or isilicitly 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:
None 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is L
not created because this change does not create the possibility of an accident or malfunction of a type l
different from those evaluated in the UFSAR.
This change has no adverse impact on system or function.
The SBLC system is tested in a manner similar to its designed j
operation and only when it'is not required to be operable.
l The testing must be completed and the system returned to an operable status before the reactor can be placed in a l
condition where the SBLC system is required.
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l SE-96-021 CONTD 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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E04-2-95-027 Standby Liquid Control Pump 2A Discharge Check Valve Replacement DESCRIPTION:
This design change replaced the 2-1101-43A lift-check valve located on the discharge of the 2A SBLC pump with a piston-check variant.
The new piston-check valve is slightly oversized for the application but is the only serviceable spare available.
The new piston-check will only open 89% of full stroke under full flow conditions.
However, its associated pressure drop is still less than the original valve.
SAFETY EVALUATION SU30tARY 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:
Anticipated Transients SAR SECTION:
For this accident, 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 SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the reactivity requirements for the SBLC system are sufficient to shutdown the reactor from full power to a cold, xenon-free shutdown in the absence of any control rod motion.
The resulting reactivity in the shutdown condition is kg S 0.97 (2 3% Ak subCritical).
A margin of 25% additional boron is added to compensate for leakage and possible imperfect mixing.
At an injection rate of 40 gal / min (one pump), the time required to inject sufficient boron to override the rate of reactivity insertion due to a cooldown of the reactor following the omenonarkmmarr..pr
E04-2-95-027 CONTD xenon peak from full rated power is approximately 83 minutes.
For an ATWS event the injection rate is 80 gal / min i
which will inject the solution in approximately 42 minutes to meet the requirements of 10CFR50.62.
Operation of the Standby Liquid Control System (SBLC) will not be detrimentally affected by installation of the new check valve.
The flow coefficient of the new valve (Cv=40) is greater than the flow coefficient for the old check valve (Cv=23) indicating that the new valve produces a smaller pressure drop than the existing valve.
The new valve does not open greater than 89% under full flow conditions and although the check valve could be sized better, the system configuration is not propitious for any check valve type.
Flow disturbances and 90* elbows within the' recommended minimum 10 pipe diameters ensure rapid wear of check valve components.
However, the new valve should equally withstand the punishment when compared to the existing valve.
It should be noted that the SBLC system is not a system which is run continuously thus the new check valves' service life may be satisfactory.
There will be no new failure modes associated with the new check valve.
Check valves can either fail open or closed.
The addition of a spring in the new valve may add another component subject to failure but the spring is constructed of stainless steel and no path exists for the spring to be j
released into the piping.
Failure of the check valve in the open position would still allow two pump flow.
Failure of
)
the check valve in the closed position would still permit i
the system to inject sodium pentaborate with one pump.
Pump operability is verified monthly using demineralized water.
If wear problems with the new check valve occur they would most likely be discovered during the monthly surveillance as the problem with the existing check valve was discovered.
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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i SE-94-032 Setpoint Change #94013I DESCRIPTION:
Setpoint of Low Pressure Emergency Core Cooling System (ECCS) Discharge Lines Pressure Switches 1(2)-1054 were changed from 45 psig (dec.) +2/-0 to 59 psig (dec.) +/-1.
SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine each accident or anticipated transient described in the j
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:
None 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the setpoint change would not create any possibility of an accident or malfunction of any type that has not been evaluated.
There is no new failure mode i
induced.
The new setpoint is more conservative, and as a result, would increase the reliability of the Low Pressure ECCS Discharge Lines Pressure Switches by decreasing the probability that the switches might actuate outside the Tech. Spec LCO limit.
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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l E04-2-95-055 Unit 2 250V Battery Charger DESCRIPTION:
The 250 VDC Battery Charger #2 feed and load breakers as well as their cables are undersized for this application.
This change replaced the 125 amp breaker at MCC 29-2, B4
)
with a 150 amp breaker and replaced the 250 amp breaker at MCC #2, A03 with a 300 amp Breaker.
SAFETY EVALUATION
SUMMARY
l 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 I
The accidents which meet these criteria are listed below:
Loss of Coolant (LOCA)
UFSAR SECTION: 15.6 For this accident, it has been determined that the change I
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 the function of the breakers is to protect the cables from overloaded conditions.
Replacing them with a proper size will not create an accident l
different from those 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.
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Setpoint Changes 96-010E & 96-011E RHR 1-1001-16A&B Thermal Overload Replacement DESCRIPTION:
The motor was replaced with a motor of hiji:er current draw.
l The motor was previously evaluated as a replacement motor l
for the RHR heat exchange bypass valves 16A and 16B.
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 l
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:
LOCA (Large Break)
SAR SECTION:
15.6 i
For this accident, 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 evaluated in the SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the operation of the MOVs have not l
changed.
This change results in a new thermal overload setting and a change in the current draw of the motor.
There are no new failure modes or system interactions.
Thus a new accident not evaluated in the SAR is not created.
3.
The margin of safety, is not defined in the basis for any Teghnical Specification, therefore, the safety margin is not reduced.
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M04-2-94-007 Core Shroud Repair DESCRIPTION:
In-vessel inspections found linear indications in the horizontal core shroud welds at Dresden Unit 3 and Quad Cites Unit 1 during the spring 1994 outages.
Visual examination and ultrasonic testing at weld H5 indicated the crack extended 360 degrees around the circumfe.ene of the shroud.
Two boat samples were taken (Quad Cit.?=.* azimuths 154 and 342-size 3" X 2" x 1.5": Dresden; azimuths 153 and 324 size 3" X 2" x 1.35") to examine / analyze the root cause of the linear indications and compare measured crack depths in the samples to the depths determined by ultrasonic testing.
Metallurgical evaluation determined intergranular stress corrosion cracking to be the root cause of the linear indications due to the. application of the welded Type 304 stainless steel components in a strongly oxidizing aqueous environment.
The depth and length of the cracking has made repairs unavoidable at these plants.
A conservative evaluation concluded that the cracked shrouds will satisfy ASME Code margins against weld failure for fifteen months of operation above cold shutdown.
The NRC approved Quad Cities Unit 1 and Dresden Unit 3 for fifteen months of operation above cold shutdown on July 15, 1994.
It is anticipated that the two online units, Dresden Unit 2 and Quad Cities Unit 2 will have sinilar linear indications and will also need repair.
The core shroud horizontal welds have a potential of failing through wall.
The technical design requirement is that the repair design structurally replaces the core shroud horizontal welds H1 through H7 if these welds fail completely through wall.
In addition, for design purposes the circumferential jet pump support plate H8 weld is to be cohdidered cracked completely through and 360 degrees.
Also, the design should not result in a driving mechanism for Intergranular Stress Corrosion Cracking (IGSCC) in these welds or any other component in the reactor vessel such that it reduces the operating margin available from the remaining ligaments of the welds.
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M04-2-94-007 CONTD The core shroud repair-is designed to structurally replace the core shroud's horizontal welds H1 through H7 and provide vertical clamping forces on the shroud in the event that any or all the seven shroud horizontal weld joints are cracked through wall.
In general the core shroud repair design j
installs low tension tie rods with spring stabilizers connected between the separator head support ring and the i
jet pump support plate.
Four tie rods will be evenly distributed in the annulus region of the reactor pressure vessel.
Spring stabilizers will be mounted at the top guide support ring (welds H2/H3) and the core plate support ring (welds H5/H6) in the annulus area between the core shroud and the reactor pressure vessel wall.
A middle spring stabilizer is mounted on the tie rod at the same elevation as the jet pump riser braces.
The upper and lower springs transmit seismic loads from the nuclear core directly to the RPV via the core plate support ring and the top guide support ring.
The function of the spring stabilizers is to provide lateral stability for the core shroud to ensure core geometry and refloodable volume are maintained.
The spring stiffness in the stabilizero was optimized to provide the minimum possible adverse effect of the seismic loads to the reactor internals (i.e. maximum horizontal support for the fuel assemblies) while meeting the stress and displacement limits.
The middle spring provides an intermediate lateral support to the tie rod and keeps the shroud from moving closer than 0.5 inches to the jet pump riser braces.
The tie rod function is to provide rotational stability for the core shroud to ensure core geometry and refloodable volume are maintained.
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 otructure, system,
)
or component could lead to the accident.
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l M04-2-94-007 CONTD I
The accidents which meet these criteria are listed below:
l Decrease in Reactor Coolant Inventory SAR SECTION:
15.6 (LOCA)
For this accident, 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 SAR.
2.
The possibility for an' accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the seismic analyses were based on the time history method of analysis.
The input motions included the 1957 Golden Gate earthquake record and a synthetic time history matching the Housner spectrum curve.
The major forces include dead load, buoyant forces, horizontal and vertical seismic, mainsteam LOCA, reactor recirculation LOCA 1
(including blowdown and acoustic), and fluid mass.
The forces were combined using the appropriate load combinations from the UFSAR.
Also considered was the combination of seismic load concurrent with each LOCA.
Analyses were done for the complete range of postulated shroud welded joint cracks as well as for the fully uncracked configuration with the shroud restraint hardware installed.
Bounding Design Basis Earth Quake (DBE) loads were obtained for use in load combinations for the Emergency and Faulted conditions, and bounding Operating Basis Earthquake (OBE) loads for the Upset condition.
The resulting seismic loads were used as input to the design of the shroud repair hardware and to validated the continued structural integrity of the core support structure and the RPV internals.
All the loads and load combinations that are relevant to the core shroud, have been evaluated and are within design allowables with the core shroud hardware in place.
The stabilizers do not add any new operational / failure mode or create any new challenge to safety-related equipment to other equipment whose failure could cause a new type of accident.
In addition, the stabilizers do not create any new component / system interactions or sequence of events that lead to a new type of accident.
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M04-2-94-007 CONTD 3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because leakage flow to bypass the steam separators due to machining eight circular holes through the jet pump support plate, cracks in the seven horizontal circumferential welds H1 through H7, cracks in the circumferential weld in the jet pump support plate H8, leakage past the jet pump support plate access hole covers, leakage paths through the shroud head flange pockets / notches, and one of the pockets / notches with a hole cut through the back of the shroud head support ring (Unit 2 only) have been evaluated.
To assure a bounding estimate, the evaluation of the bypass flow leakage is conservatively assumed that each of the shroud welds develops a complete circumerential crack gap of one mil.
These leakage flows are based on applicable loss coefficients and reactor internal pressure differences across the applicable shroud components.
The performance impact of the total bypass leakage flow for 100% rated power and 87 to 108% rated core flow is discussed below:
CORE MONITORING:
Measured " total core flow" (actually cumulative flow through l
the pumps) is an input to the core monitoring computer code's power distribution calculation.
These are performed at least daily during steady-state operation above 25% power to demonstrate compliance with the core operating limits as required by Technical Specifications.
The code adjusts (reduces) this measured total jet. pump flow to account for flow that does not pass active fuel rods (i.e. Ex-Channel and water rod flow). The ex-channel bypass flow does not account for the new potential leakage paths associated with the shroud.
A conservative estimate on the ir. pact from the I
various shroud leakage paths on these calculations is an l
indicated active core flow that is about 0.22% higher than actual.
This is small compared to.the core flow measurement uncertainty of 2.5% for jet pump plants (Reference 1) used in the uncert..nty analysis associated with the Minimum I
l Critical Power Ratio (MCPR) Safety Limit.
Additionally, the affect of having 0.22% lower core flow than indicated by the core monitoring code is only a 0.1% decrease in MCPR relative to that calculated during these surveillances.
Because this small difference only affects operating margin (margin at steady-state compared to the MCPR operating limit), the margin of safety is not affected.
The effect on other core surveillance parameters (LHGR and MAPLHGR) would be even smaller and also insignificant.
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M04-2-94-007 CONTD FUEL THERMAL MARGIN EFFECT - ANTICIPATED ABNORMAL I
The code used to evaluate performance under anticipated abnormal transients and determine fuel thermal margin includes carry under as one of the inputs.
The effect of the increased carry under due to leakage results in greater compressibility of the downcomer region and, hence, a reduced maximum vessel pressure.
Since this is a favorable effect, the thermal limits are not impacted.
EMERGENCY CORE COOLING SYSTEM (ECCS):
The leakage flow above the top guide support ring results in slightly increased carry under that causes the initial core enthalpy to increase slightly, with a corresponding decrease in the core' inlet subcooling.
However, because the total downconer carry under still meets the design value, there is no impact on the ECCS performance from this condition.
1 Another effect of the leakage flows from the repair holes and the weld cracks is to decrease the time to core uncovery slightly and, also to increase the time that the core is uncovered.
The combined effect has been assessed to increase the Peak Clad Temperature (PCT) for the limiting LOCA event (Reference 2) by less than 15 degrees F.
The current analysis basis yields LOCA PCTs of approximately l
1680 degrees F for the diesel generator failure case.
Therefore substantial margin exists to the 10CFR50.46 acceptance criterion of 2200 degrees F.
Because the maximum potential effect on the design basis LOCA PCT is very small, there is no adverse effect on the margin of safety.
This impact is sufficiently small to be judged insignificant, and, hence, the licensing basis PCT for the normal condition with no shroud leakage is applicable.
The sequence of events remains essentially unchanged for the LOCA events with the shroud head leakage.
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2 UPDATE QUAD CITIES UFSAR r
DESCRIPTION:
The purpose of this change is to update the Quad Cities UFSAR, due to the use of the NFS approved, NRC licensed GE8X8NB-3 (GE10) fuel design currently used in the Q1C14 and Q2C14 Reloads.
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:
i Loss of Feedwater Heating FSAR SECTION:
15.1.1 Feedwater Controller Failure FSAR SECTION:
15.1.2 Load Rejection with No Bypass F3AR SECTION:
15.2.2 Turbine Trip with No Bypass FSAR SECTION:
15.2.3 ASME Overpressurization Event FSAR SECTION:
15.2.4 Transients During SLO FSAR SECTION:
15.3.6 Rod Withdrawal Error FSAR SECTION:
15.4.2 Misoriented Fuel Assemblies FSAR SECTION:
15.4.8 Control Rod Drop Accident FSAR SECTION:
15.4.10 DBA LOCA FSAR SECTION:
15.6 Design Basis Fuel Handling FSAR SECTION:
15.7.2 Accident Reactor FSAR SECTION:
4 l
New Fuel Storage FSAR SECTION:
9.1.1 l
Spent Fuel Storage Racks FSAR SECTION:
9.1.2.2.3.2 For 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 l
malfunction of equipment important to safety as previously i
evaluated in the FSAR.
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f UPDATE QUAD CITIES UFSAR CONTD 2.
The possibility for an accident or malfunction of a-I different type than any previously evaluated in the UFSAR is not created because adequate margins have been shown in the reload analyses to fuel thermal-mechanical design limits.
The GE8x8NB-3 design is neutronically compatible with the existing fuel types and core components in the Q1C14 and i
Q2C14 cores.
The NRC has reviewed and approved the GE8x8NB-3 fuel design and this approval is documented in General Electric Standard Application for Reactor Fuel II (GESTAR
)
II) NEDE-24011-P-A-10, Amendment 21.
The GE8x8NB-3 bundle is hydraulically identical to the l
GE8x8NB bundle with the exception of the flow trippers on i
l the upper one third of the. channel which are designed to improve margin to CPR.
The bundle pressure drop characteristics of the GE8x8NB-3 bundle are identical to i
those of the GE9 fuel design, hence core thermal-hydraulic stability characteristics are not adversely affected by the GE8x8NB-3 design.
The GE8x8NB-3 design does not adversely impact equipment important to safety and does not create the possibility of a l
new or different kind of accident scenario different from those previously evaluated in the FSAR.
]
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3.
1.1.A - Fuel Cladding Integrity Safety Limit (MCPR Safety Limit) i 3.5.K - Minimum Critical Power Ratio I
i The margin to safety is not reduced since the Safety Limit MCPR (SLMCPR) in the Quad Cities Technical Specifications (1.07) is adequate to ensure that 99.9% of the rods will not experience boiling transition during an anticipated transient when the MCPR limit in Section 3.5.K is maintained or exceeded.
The GE10 fuel design has been generically analyzed with approved methods per GESTARII NEDE-24011-P-A-10, Amendment 21 and the use of the 1.07 Safety Limit MCPR i
value has been previously approved as conservative for application to GE10 fuel in D-lattice plants such as Quad Cities.
The GE8x8NB-3 channels contain flow trippers which are designed to disrupt the liquid flow on the channel wall, l
divert the flow onto the fuel rods,.and thereby improve the l
MCPR performance of the bundle.
The MCPR improvement is incorporated into the calculation of the'R-factor for the bundle which is used by CMC to convert the bundle power to CRR.
Therefore the margin of safety is not reduced with the
(
use of the generically approved GE8x8NB-3 fuel.
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l UPDATE QUAD CITIES UFSAR CONTD 1.1.c - Fuel Cladding Integrity Safety Limit (Power l
Transient) t The fuel thermal time constant for 8x8 arrays is approximately 5.0 seconds.
Both GE8X8NB (GE9B) and GE8x8NB-i 3 (GE10B) designs have 8x8 arrays.
The margin of safety for prctecting the cladding from heat flux is not reduced as a result of the GE8x8NB-3 fuel design.
The GE10 fuel design has been generically analyzed with approved methods per
(
GESTAR II NEDE-24011-P-A-10, Amendment 21 for use in D-lattice BWR 3/4's.
1.2 - Reactor Coolant System Safety Limit The margin of safety is not reduced as a result of the GE8x8NB-3 design used as the Quad Cities Reload fuel.
With i
respect to-overpressurization transients, the GE8x8NB-3 is identical to the previously approved GE8x8NB fuel.
The GE8x8NB-3 fuel design has approximately the same fuel l
thermal fuel time constant as the GE8x8NB fuel.
The GE10 l
fuel design has been generically analyzed with approved methods per GESTARII NEDE-24011-P-A-10, Amendment 21 and the use of the 1.07 Safety Limit MCPR value has been previously approved.
The GE8x8NB-3 fuel reloads are evaluated by NRC licensed methods each cycle via the Supplemental' Reload Licensing Report to ensure that adequate margin to the vessel pressure limit is maintained.
3.5.I - Average Planar LHGR i
3.5.J - Local LHGR The margin of safety for the Average Planar LHGR and Local LHGR is not reduced as a result of using the GEx8xNB-3 design at Quad Cities Station.
The GE8x8NB-3 is very similar to the previously approved GE8x8NB fuel with respect to Average Planar LHGR and Local LHGR, hence the local or nodal power distributions will remain approximately the same as the GE8x8NB design.
The margin of safety for protection i
against: 1) 1% plastic strain, 2) centerline melt, and 3) initial energy assumed in the LOCA calculations is not reduced with the use of GE8x8NB-3 at Quad Cities Units 1 and 2.
The GE8x8NB-3 fuel design has b-". generically analyzed with NRC approved methods per GESTAh II NEDE-24011-P-A-10, Amendment 21.
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l M04-1-88-103C DESCRIPTION:
l The RHRS large bore piping was evaluate for a 340 degree F l
shutdown cooling mode temperature.
The piping modifications f
brings the piping system within code design margins.
BAFETY EVALUATION
SUMMARY
l l
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.
I i
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
l Operation or failure of the changed structure, system, I
or component could lead to the accident.
(
The accidents which meet these criteria are listed below:
5.6.5.2 For this accident, 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 the failure of the lines and the consequences of the failure have decreased due to this modification.
The RHR piping lines are being resupported to meet the FSAR and Mark I loading requirements.
Based on this, this change does not adversely impact systems or functions so as to create the possibility of an accident or l
malfunction of a different type from those evaluated in the SAR.
l 3.
The margin of safety, is not defined in the basis for any j
Technical Specification, therefore, the safety margin is not reduced.
l l
I' Q:\\MiCHOP3\\SAFFIYuMRPT wpf
M04-1-88-103A 1
l i
l l
DESCRIPTION:
RHRS large core piping was evaluated for a 340' F shutdown l
cooling mode temperature.
The following modifications i
brings the piping system within code design margins.
j Revising / Adding / Removing thirty-eight (38) pipe supports.
In addition, tee reinforcement and reinforcing weld.
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:
l 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:
15.6.5.2 SAR SECTIONS:
14.2.4 For this accident, 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 the failure of the piping and supports and the consequences of.their failure are the same as they were prior to the modifications.
Based on this, this change does not adversely impact systems or functions so as to create the possibility of an accident or malfunctions of a l
different type from those evaluated in the SAR.
3.
The margin or safety, is not defined in the basis for any j
Technical Specification, therefore, the safety margin is not l
reduced.
i Q:\\TECHOPMSAFETY\\l96RFT.wpf
I i
M04-1-91-027B DESCRIPTION:
This partial modification consists of the replacement of Intergranular Stress Corrosion Cracking (IGSCC) susceptible piping, valves, Regenerative Heat Exchangers (RHX) and e
l instrumentation in the non-safety related portion of the l
" Reactor Water Cleanup" system.
The piping and equipment i
are being replaced with piping'and equipment fabricated from SA 312 Type 316/316L stainless steel with special. chemistry
)
restrictions of maximum carbon content of 0.020% and a maximum cobalt content of 0.10%.
This material is highly resistant to IGSCC.
e l
This partial modification includes replacement of the piping that runs from the outboard side of the MOV 1-1201-5 valve up to the RHX, the interconnecting piping between the RHX i
and non-regenerative heat exchangers (NRHX), sections of the pump suction piping from the 1-1201-148A/B valves to the
- pumps, the pump discharge piping from the pumps to the 1-1299-11 valve.
Then from a point upstream of the shell side i
of the RHX to the 1-1201-82 valve.
The following additional design enhancements / changes were performed:
1.
The carbon content for the replacement stainless steel material was limited to provide a high degree of resistance to IGSCC.
The cobalt content was limited to reduce the amount of cobalt in the primary system.
All the single disc _ gate valves in the main process lines were replaced with double-disc gate valves and will have Stellite free hardfacing to reduce the' amount of cobalt.
These valves provide positive isolation characteristics and are not susceptible to thermal i
binding.
The vent and drain valves were replaced with quick maintenance type valves.
The electropolished piping and pipe bends were used instead ot fittings wherever possible to reduce radiation, buildup on the piping.
The RHX channels are electropolished to reduce radiation buildup.
The new RHX has the strip baffle design which will prevent crud buildup by keeping solids in suspension and avoiding any abrupt direction changes.
emenomsasmmarr..,r
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M04-1-91-027B CONTD 2.
The existing dual train heat exchanger.(HX) design was replaced with a single train HX design.
The two existing stacks of RHXs were replaced with one stack of three shells connected in series.. The new single HX stack is cag :ble of the same heat removal capacity and flow rate of both existing RHX stacks running in parallel to the existing NRHX stacks.
A single isolation valve is provided upstream and downstream of each NRHX stack to provide isolation of a NRHX train.
This single train arrangement is designed for a total flow rate of 2% of Feedwater flow (410 gpm).
The maximum low rate capacity is the same as the current system when both trains are operated in parallel at full flow.
The function of the RWCU system will not j
change.
The station can operate the system from 1% to i
2%,-which is the same as the original system design.
1 The single train configuration has the advantage of reducing the amount of equipment in the system.
Sixteen process valves and forty eight RWCU vent and drain valves will be eliminated.
This will reduce future maintenance and radiation crud traps on the system.
3.
A RHX bypass line was added so that the RWCU system can be used as an alternate shutdown cooling system when the reactor is in shutdown and the. vessel temperatures are less than or equal to 200 degrees Fahrenheit.
4.
RWCU_ pump hot suction cross-connect piping 1-12121-4"-
A, 1-12123-4"-A; and cross-connect valves 1-1299-9, 1-1299-10 and 1-1299-12 were permanently removed so that the RWCU pumps cannot be cross-connected to the high temperature side of the system.
Experience has shown that operating the pumps on the high temperatura side has caused a high failure rate of the pump mechanical l
seals.
The cross-connect piping 1-12124-4"=A will be i
maintained for use as a reverse flow path for the system chemical decons.
This piping has never seen high temperatures.
5.
The instrumentation was replaced and modified to match j
the single train configuration.
New flow indicators FI 1-1201-173A/B were added to provide local flow i
i indication in gpm.-
Dp switches DPIS 1-1201-174A/B were i
relabeled as FS 1-1201-173A/B.
The pump minimum trip set point for FS 1-1201-173A/B was increased from 30 i
gpm to 55 pgm.
The set point increase is in the l-conservative direction.
1 Q:\\TECHOPESAFETY\\l%RFT.wpf
1 M04-1-91-027B CONTD TS 1-1291-8 which provides a high temperature alarm to the control room was eliminated.
TIS 1-1291-13 was replaced with a thermocouple input indicator dual switch unit to provide the high temperature alarm to the control room that was supplied by TS 1-1291-8 and provide an isolation signal to the RWCU isolation valves which is the function of existing TIS 1-1291-13.
Existing flow elements orifices FE 1-1279-74A/B were increased.
The flow controller FC 1-1279-97A/B were replaced and an increase from 2" to 3" diameter valve seats for the Demin. Inlet Isolation valves AO 1-1279-14A/B.
Flow Recorders FR 1-1279-97A/B were eliminated.
The increased flow capability necessitates an increase in the range of the
'A' and
'B' Demin. Flow indicators in the control room.
Also,'the flow indicators FI 1-1290-30A/B scale was increased.from 0-150 gpm to 0-250 t
gpm.
6.
The reactor pressure vessel preheat system was permanently disconnected from the RWCU system since it has never been used to preheat the RWCU system.
7.
The piping layout and valve locations changed to accommodate the single train design, optimize pipe runs
+
lengths, improve accessibility, and reduce radiation exposure for construction, operation, and maintenance.
A gallery platform was added to the RWCU HX Room and two penetrations #131 and #43 for secondary containment and fire p'rotection were abandoned and relocated.
8.
The shield wall plugs for the RHX were permanently removed since the tube bundles for the new HXs are non-removable.
9.
Three abandoned 3/8" O.D.
copper tubing sample lines 1-I 8804A,B and C that were part of the Primary Containment Sample System (PCSS) that runs through the RWCU HX room was removed. -These lines are interferences and are no longer needed.
They were to be removed under Modification M04-1-92-011.
10.
The replacement RHX was designed and fabricated to ASME Section,4II, 1992 Edition with the supplemental requirements of ASME Section III 1965 Edition and the requirements of the original General Electric RHX I
design specifications.
l Q:\\TBCHOPS\\S/JTTY\\l96RFT.wyf
M04-1-91-027B CONTD
{
11.
Control and power cables to RWCU return line isolation i
valve MOV 1-1201-80 was re-classified as non-safety related.
The valve is classified as a non-safety related valve.
12.
A local leak rate test tap was added to the piping just l
downstream of the MOV 1-1201-5 valve to allow local j
leak rate testing of the RWCU Primary Containment Isolation valves.
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:
Postulated Piping Failures in Fluid Systems Outside Primary Containment.
SAR SECTION:
3.6.1 For this accident, 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 SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because this partial modification is to replace the non-safety related portion of the RWCU system including the Regenerative Heat Exchanger and the associated piping, l
instruments and valves.
These components are being replaced with a special stainless steel which reduces the possibility l
of IGSCC.
The changes in the piping configuration and size have been evaluated and approved.
This Non-Safety Related l
portion of the RWCU system will not adversely affect the Safety Related portion starting at the MO 1-1201-05 valve.
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M04-1-91-027B CONTD l
The Safety Related valve has been protected from the RWCU system with an existing anchor support and as such the modification will not create a scenario where the MO 1-1201-05 valves is adversely affected.
There will not be additional interfaces with other Safety Related systems and thus this Partial Modification will not impact additional components and create the possibility of an accident or malfunction different than those currently analyzed.
3.
The margin.of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
1 I
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1 1
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- ~
E04-1-93-323 DESCRIPTION:
Valves MO-1-1402-24A(B) are the outboard injection valves for the Core Spray System.
These motor operated valves provide containment isolation as well as the capability for isolating the Core Spray System discharge header for maintenance or testing.
The valves are normally open.
The MO-1-1402-24A(B) valves are interlocked such that they can be opened only if the associated inboard injection valve (M01-1402-25A(B)) is in the closed position.
This interlock is bypas'ed when the reactor pressure is less than 325 psig.
s Automatic opening of the outboard injection valves can only be accomplished when an initiation signal is present and reactor pressure is less than 325 psig.
This Exempt Change increased the thrust output of the valve actuators to ensure the valve can be positioned against system design conditions.
This Exempt Change also increased the Structural Thrust Limit of the valves by replacing limiting components with new components having higher thrust limits.
To increase the valve actuators thrust output the overall gear ratio of the actuator was changed from 29.44 to 59.4 and the springpacks were changed to a No. 0901-211.
Additionally, the existing 3600 rpm, 60 ft-lb actuator motors were replaced with a new 1800 rpm, 80 ft-lb motors.
The impact of the new motors on the electrical system was reviewed by Bechtel.
The existing Thermal Overload Heater was replaced to provide proper protection for the motors (Bechtel calc. No.: QC-470-E-001, Rev. 0).
The existing power cables were upgraded (Bechtel Calc. No.: QC-014-E-004, Rev. 0).
The existing circuit breakers are adequate for the new motors (Bechtel Calc. No. QC-442-E-002, Rev. 5).
These i
changes increased the valves stroke time from 9.9 seconds to 40 seconds.
To increase the valves Structural Thrust Limit the existing disk was replaced with new disk higher structural thrust limits.
1 I
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E04-1-93-323 CONTD 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 Coolant Accidents Resulting from Piping Breaks Inside Containment SAR SECTION:
15.6.5 For this accident, 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 SAR.
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 failure modes or system interfaces are created.
These changes will not modify the function of the valves, nor will they affect any interactions with other safety related componcnts or systems.
The impact of the new motor on the electrical distribution system was reviewed.
The existing Thermal Overload Heater will be replaced to provide proper protection for the motor (Bechtel Calc. No.: QC-470-E-001, Rev. 0).
The existing power cables will be upgraded (Bechtel Calc. No.: QC-014-E-004, Rev. 0).
The existing circuit breakers are adequate for the new motors (Bechtel Calc. No.: QC-442-E-002, Rev. 5).
The Core Spray System piping qualifications were updated to include the weight change due to the new components and is acceptable (VECTRA I
Calc. Nos.: 28.0201.1033.013, Rev. 0; 28.0201.1033.134, Rev.
j 0; and 28.0201.1033.02, Rev. 2) Therefore these changes will j
not create a new accident scenario or malfunction not already evaluated in the UFSAR.
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E04-1-93-323 CONTD l
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the margin i
of safety is not reduced, because the Core Spray subsystem will still provide the flow to the reactor core within the time assumed in the accident analysis.
Previous analysis l
assumed that the injection valves would pass no flow until l
the valves had stroked full open.
The new analysis l
accurately models the flow through the gate valve as it strokes open.
By slowing down the valve, the valve is made i
more reliable in terms of the thrust generated by the l
operator.
Slowing down the valve enables compliance with the NRC Generic Letter (GL) 89-10 MOV guidelines.
I l
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i
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l E04-1-93-324 A
i DESCRIPTION:
Valves MO-1-1402-25A(B) are the outboard injection valves for the Core Spray System.
These motor operated valves provide containment isolation as well as normally isolate the Core Spray System from the reactor vessel.
The valve is automatically opened when the Core Spray System is initiated and reactor pressure becomes less than 325 psig.
The valve can be throttled to control system flow below 4500 pgm.
The MO-1-1402-25A(B) valves are interlocked in the closed position when reactor pressure is greater than 325 psig.
Additionally, the valves are interlocked with their associated outboard injection valves [MO-1-1402-24A(B)) to prevent opening both injection valves simultaneously at high j
reactor pressures.
This Exempt Change increased the thrust output of the valve actuators to ensure the valve can be positioned against i
system design conditions.
This Exempt Change also increased the Structural Thrust Limit of the valves by replacing j
limiting components with new components having higher thrust limits.
To increase the valve actuators thrust output the overall gear ratio of the actuator was changed from 29.44 to 59.4 l
and the springpacks were changed to a No. 0901-211.
Additionally, the existing 3600 rpm, 60 ft-lb actuator noters were replaced with a new 1800 rpm, 80 ft-lb motors.
The impact of the new motors on the electrical system was reviewed by Bechtel.
The existing Thermal Overload Heater was replaced to provide proper. protection for'the motors l
(Bechtel Calc. No.: QC-470-E-001, Rev. 0).
The existing l
power cables were upgraded (Bechtel Calc. No.: QC-014-E-004, Rev. 0).
The existing circuit breakers are adequate for the new motors (Bechtel Calc. No. QC-442-E-002, Rev. 5).
These changes increased the valves stroke time from 9.9 seconds to 40 seconds.
To increase the valves Structural Thrust Limit the existing disk and stem were replaced with new components with higher structural thrust limits.
)
i j
Q:\\1ECHOP3\\SAFEIY\\l96RPT.wpf l
l
-l 1
E04-1-93-324 CONTD SAFETY EVALUATION
SUMMARY
1.
The change described above has been analyzed to determine l
each accident or anticipated transient described in the UFSAR where any of the following is true:
l The change alters the initial conditions used in the i
UFSAR analysis.
I I
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accidenc.
Operation or failure of the changed structure, system, l
or component could lead to the accident.
The accidents which meet these criteria are listed below:
Loss of Coolant Accidents Resulting from Piping Breaks Inside Containment SAR SECTION:
15.6.5 For this accident, 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 SAR.
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 failure modes or system interfaces are created.
These changes will not modify the function of the valves, nor will they. affect any interactions with other safety related components or systems.
The impact of the new motor on the electrical distribution system was reviewed.
The existing Thermal Overload Heater will be replaced to provide proper protection for the motor (Bechtel Calc. No.: QC-470-E-001, Rev. 0).
The existing power cables will be upgraded (Bechtel Calc. No.: QC-014-E-004, Rev. 0).
The existing circuit breakers are adequate for the new motors (Bechtel Calc. No.: QC-442-E-002, Rev. 5).
The Core Spray System piping qualifications were updated to include the weight l
change due to the new components and is acceptable (VECTRA Calc. Nos.: 28.0201.1033.013, Rev. 0; 28.0201.1033.134, Rev.
0; and 28.0201.1033.02, Rev. 2) Therefore these changes will l
not create a new accident scenario or Falfunction not I
already evaluated in the UFSAR.
L Q3TECHOP3\\ SAFELY \\l96RIT.wpf
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1 E04-1-93-324 CONTD 3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the margin i
of safety is not reduced, because the Core Spray subsystem l
will still provide the flow to the reactor core within the time assumed in the accident analysis.
Previous analysis i
l assumed that the injection valves would pass no flow until the valves had stroked full open.
The new analysis accurately models the flow through the gate valve as it opens.
By slowing the valve down, the valve is made more reliable in terms of the thrust generated by the operator, t
Slowing down the valve enables compliance with the NRC Generic Letter (GL) 89-10 MOV guidelines.
I 1
I l
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l I
l 4
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l l
QCFHP 100-1 DESCRIPTION:
Added a prerequisite to pull the spent fuel pool transfer canal gates (SFPTCG) prior to fuel movement.
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 4
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 SFP inventory UFSAR SECTION: 3.5 For this accident, 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 the procedure change will allow QCNPS to perform refueling operation with the SFPTCG out.
The loss of fuel pool inventory is a evaluated accident or malfunction in the SER.
The effect of this change would be that inventory would be lost from both pools, instead of one.
This is not a new or different type of event (loss of inventory).
The conservative time to boil assumptions would now increase because the volume of water has doubled, which would slow the rate of level loss from the rate if only one pool provided the inventory.
3.
The margin of safety, as defined in the basis for any t
Technical Specification, is not reduced because the margin of safety is not reduced, because the Tech Spec limit of 33 feet is based upon allowing enough water level to shield and cool the fuel assemblies. The change does not affect this requirement, as the Tech Spec applies to both fuel pools.
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SE-96-038 FIF 96-0960 DESCRIPTION:
G.E.
P.O. No. 205-H0386 for the RHR Pumps is dated 6-7-67,.
G.E. Purchase Spec 21A5791.for RHR Pumps, Rev.
5, (equivalent to a Code Design Spec) is dated 1-11-69 (and specifies class C).
The original SER (dated 8-25-71) reads
" Pumps for the RHRS conform to the Class C requirements of Section III".
Also note that the UFSAR stipulates Class C for the RHR HX's which are in the same system and should be the same classification.
Per ASME Code,Section III, 1965 through 1970, the RHR Pumps would be classified Class C (Ref. Paras. N-132 and N-133) in the time frame of the pumps' purchase.
Reg. Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, Radioactive-Waste-Containing Components of Nuclear Power Plants", Ref. 3 dated F~bruary 1976, Paragraph C.2.a stipulates that Group C standards should be applied for post-accident heat removal and residual heat removal (e.i., RHR).
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:
l Decrease in Heat Removal by the l
Reactor Coolant System UFSAR SECTION: 15.2 Decrease in Reactor Coolant Inventory UFSAR SECTION: 15.6 Anticipated Transients Without Scram UFSAR SECTION: 15.8 For thase 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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J SE-96-038 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 RHR Pumps were specified, purchased and built to ASME Class C requirements.
This change is a UFSAR only change, bringing the UFSAR into agreement with these documents as well as the August 25, 1971 SER.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the change of UFSAR ASME Class designation does not affect the equipment functional capability and, hence, does not affect Tech Spec limits.
1 i
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1 l
M04-1-95-006 Scram Discharge Volume l
DESCRIPTION:
A modification was required to restore the Scram Discharge i
Volume (SDV) logic to the requirements outlined in IEEE 279-68 single failure criteria.
Two new junction boxes were mounted near the existing junction boxes on both the north and south Scram Discharge Volumes (SDV).
The RIS level switches were removed from the existing boxes and mounted in the new junction boxes.
Conduit and cables were reworked as required to support the relocation.
For the north side the new boxes supply sub-channels Al and Bl.
The south side r
junction boxes provide trip inputs to subchannels A2 and B2.
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, i
The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident.
i Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
15.8 For this accident, 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 j
evaluated in the SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the' change will allow the equipment to function as it was intended per the UFSAR.
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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M04-1-95-007 Core Spray T-Box Repair Clamp DESCRIPTION:
This modification involved the portion of the Core Spray "B"
)
loop system located inside the Reactor Pressure Vessel (RPV) annulus of Quad Cities Unit 1.
The piping enters the annulus through RPV penetration (i.e. thermal sleeves) located at Azimuth 185, 472 inches above Reactor vessel zero.
The RPV annulus portion of the piping consists of a semicircular loop fabricated from type 304 stainless steel.
The core spray junction or T-box located in the annulus region at the thermal sleeve penetrations, directs core spray system flow to the attached lateral distribution piping.
The lateral distribution piping then feeds the core spray sparger through two core shroud penetrations.
1 The function of the modification is to assure the structural and geometric integrity of the core spray line in the event the reported defect were to grow to the full circumference of the pipe.
The repair utilizes an external clamp, mechanically attached to the Core Spray (CS) line laterals i
on either side of the T-Box at Azimuth 185.
The repair j
clamp _ design has four individual bolts which extend across the T-Box.
The bolts are tensioned to secure the individual clamp halves on either side of the T-box and to provide 1
frictional resistance across the crack surface under normal j
operating conditions.
Under postulated LOCA conditions, i
during the initial CS injection, thermal contraction of the CS line relative to the repair clamp may cause a loss of bolt preload.
This potential loss of reload has been addressed and the resulting clamp component stress levels
)
have been shown to be acceptable.
\\
All threaded fasteners are fitted with lock welded retainers.
These retainers (keepers) shall be tack welded in place.
The tack welds are non stressed, non structural i
The design intent of the tack welded keepers is to protect the threaded portion of the bolts and to prevent the keepers from loosening.
The tack welds for the keepers are considered non Safety Related.
The repair ensures that the CS laterals perform their intended design function of supplying Core Spray Flow to the spargers located inside the Core Shroud.
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I l
l M04-1-95-007 l
l 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.
I 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:
15.6.5 For this accident, it has been determined that the change described above will not increase the probability of an i
occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is j
l not created because all the loads and load combinations.that I
are relevant to the core spray line, have been evaluated and are within design allowables with the core T-box hardware in place.
The T-box does not add any new operational / failure mode or create any new challenge to safety-related equipment i
or other equipment whose failure could cause a new type of i
l accident.
In addition, the T-box does not create any new 1
l component / system interactions or sequence of events that j
lead to a new type of accident.
The function of the modification is to assure the structural integrity of the core spray line even if the reported defect were to grow to l
the full circumference of the pipe.
i The repaired core spray components with crack indication impact the results of some of the loss of coolant accidents.
These components are passive items inside the reactor vessel.
All accidents considered in the UFSAR are those which may result in radiological doses to the public, such as LOCA, positive reactivity insertion, and handling of j
radioactive materials inside or outside the containment.
However, any postulated accident caused by the repaired j
passive reactor internal components do not result in radiological doses to the public at the site boundary worse than the accidents already covered in the UFSAR.
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M04-1-95-007 CONTD The potential opening of a crack in the core spray line is limited on account of the actual repair clamp geometry, and is also bounded by ECCS analysis results.
Because of the repair clamp on the core spray line, the function of the i
core spray is restored.
Consequently, the geometrical integrity of the core spray line and T-box area is stabilized.
Thus there is no adverse impact to systems or function so as to create the possibility of an accident or malfunction different than those described in the UFSAR.
l 3.
The margin of safety, as defined in the basis for any l
Technical Specification, is not reduced because there are existing leakage locations in the CS line in addition to those associated with the reported crack.
Calculated leak flows from these, considering the initial LOCA temperature differentials have been evaluated.
This leakage would go into the reactor, outside the shroud, and would therefore, not contribute to cooling and reflooding of the core in a recirculation line LOCA accident.
The minimum required system capability is 4,700 gpm with the reactor at 90 psig based on the core spray system acceptance criteria specified in the station surveillance procedures, QCOS 1400-01 and 1400-4..These station surveillance procedures account for a documented 200 gpm CS system leakage.
Previous LOCA analyses assumed 4,500 gpm core spray. delivery inside the shroud when the reactor is at 90 psig relative to the drywell, therefore, an excess system capability of 200 gpm exists for leakage.
The total leakage for the T-box crack at maximum separation, the existing leakage, and the repair is 277 gpm.
This exceeds the documented 200 gpm excess system capability and thus impacts the existing LOCA analyses.
The potential for further degradation (cracking) in the CS system still exists.
Thus, the analysis added i
additional leakage beyond the 277 gpm total leakage evaluated for this modification to provided design l
flexibility (leakage margin) to be used in the future, if needed.
Hence, a bounding leakage of 400 gpm was evaluated on the ECCS performance, resulting in a CS flow of 4100 gpm injected inside the shroud compared with the rated CS flow of 4500 gpm for one loop at vessel pressure of 90 psig.
The repair design also considered the effects of relative i
thermal expansion between the repair hardware and CS piping i
in both design bases and beyond~ design basis loading i
conditions.
The impact has.been assessed and indicates an i
increase in the fuel's peak cladding temperature (PCT) of l
less than 40' F compared with the current licensing basis.
This assessment is based on the SAFER /GESTR analysis performed to support the safety evaluation of the reactor omcuan\\sAremmnn..pt
M04-1-95-007 CONTD internals configuration for the 1994 QC 1 restart.
The current loss-of-coolant-accident (LOCA) analysis basis yields a bounding licensing PCT of 1725'F for.the desigr; basis LOCA event (through the end of Cycle 15).
The 10CFR50.46 regulatory limit PCT is 2200*F.
The maximum potential isopact of the CS leakage on the design basis LOCA l
PCT is small (< 40*F) compared with the regulatory limit, thus the applicable ECCS regulatory requirements are met.
The sequence of events remains essentially unchanged for the i
LOCA events with the CS leakage, and long-term cooling is not affected.
However, the CS repair PCT evaluation, along j
with other previous PCT penalties from part evaluations, meet the 50*F threshold for reporting within thirty days as specified in 10CFR50.46.
Comed will initiate notification to the NRC per 10CFR50.4C within thirty days.
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4 DESCRIPTION:
E04-1-93-198 installed new hydraulic actuators on feedwater regulating valves (FRV) 1-0642A/B, A new fire protection 4
i system was installed to mitigate the addition of a large amount of combustible material (approximately 100 galls of l
hydraulic oil).
A new drainage and curbing barrier system will remove water during a suppression system actuation and j
remove oil should a pipe rupture.
Installation ond testing activities were evaluated.
j INSTALLATION / TESTING:
l Final installation was performed with the feedwater regulating valves out of service and isolated from 3
operational plant equipment.
Equipment may be installed and i
connections made, but final installation of the new 4
actuators and construction / modification testing was l
performed with the valves isolated.
The fire protection system was installed with the fire system in service.
A hot tap was performed to. breach the i
system.
An isolation valve was installed and maintained j
closed during installation.
Electrical isolation should be maintained during installation.
During testing, the new j
fire protection system were electrica]ly and mechanically i
returned to service as required by approved testing j
procedures.
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.ccident.
The accidents which meet these criteria Tre listed below:
Increase in Feedwater Flow SAR SECTION:
15.1.2 Loss of'Feedwater SAR SECTION:
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E04-1-93-198 CONTD For these accidents, it has been determined that the change described above vill 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 UFSAv.
2.
The possibility for ru accident or malfunction of a different type than.
,f previously evaluated in the UFSAR is not created because the operation of the new valve operators will be identical to the existing operators.
The new operators are more reliable and accurate.
The only changes that could adversely affect other equipment is the addition i
of 100 gallons of hydraulic oil (fire hazard).
The addition of the fire hazard has been mitigated by installing a fire detection / suppression system.
The new valve operators are designed to lock in place upon the detection of any operational errors, which is consistent with the existing operators.
No new accidents or malfunctions have been identified that have not been evaluated.
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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l M04-1-91-027A RWCU - Demolition of Existing Regenerative Heat l
DESCRIPTION:
M04-1-91-027 Partial A is the modification to remove the l
RWCU Regenerative Heat Exchangers and their associated piping, valves and instrumentation.
These components are being removed to eliminate the potential for IGSCC by l
replacing the components with SA 312 type 316/316L stainless l
steel with a maximum carbon and cobalt content of 0.020% and 0.10% respectively.
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:
Postulated Piping Failures in Fluid Systems Outside Primary Contefrennt SAR SECTION:
3.6.1 For this accident, 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 SAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the RWCU System is non-safety related.
The demolition partial will merely remove the regenerative heat exchangers and their associated piping, valves and instruments.
The containment isolation valve MOV 1-1201-5 will be closed per an OOS and it's function maintained.
This will not create the possibility of an accident or malfunction different than 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.
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