ML20236Q278
| ML20236Q278 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/13/1998 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20236Q277 | List: |
| References | |
| NUDOCS 9807200133 | |
| Download: ML20236Q278 (14) | |
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i PROPOSED ANO-2 TECHNICAL SPECIFICATION BASES CHANGES l
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N 9907200133 990713 l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence I
that the specified acceptable fuel design limits (i. e., DNBR and centerline fuel melt temperature) are not exceeded during normal operation and design basis anticipated operational occurrences.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III of the ASME Code for Nuclear Power Plant Components. The reactor vessel, steam generators and pressurizer are designed to the 1968 Edition, Summer 1970 Addenda; piping to the 1971 Edition, original issue; and the valves to the 1968 Edition, Winter 1970 Addenda"'.
Section III of this code permits a l
maximum transient pressure of 110% (2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
I The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety Limits during norum1 operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.25 and 21.0 kw/ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to i
trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.
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"3 Use of a later ASME Section III code is acceptable, provided the code section(s) is reconciled.
l ARKANSAS - UNIT 2 B 2-2 Amendment No. M. M. ~J-9, MS,.
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES l
Pressurizer Pressure-High i
The Pressurizer Pressure-High trip, in conjunction with the l
pressurizer safety valves and main steam safety valves, provides reactor i
coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at s 2370.887 psia which is below the nominal lift setting (2500 psia) of the pressurizer
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safety valves and its operation avoids the undesirable operation of the l
pressurizer safety valves.
Pressurizer Pressure-Low The Pressurizer Pressure-Low trip is provided to trip the reactor and 3
to assist the Engineered Safety Features System in the event of a Loss of l
Coolant Accident.
During normal operation, this trip's setpoint is set at l
2 1686.3 psia. This trip's setpoint may be manually decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at s 200 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.
Containment Pressure-High
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The Containment Pressure-High trip provides assurance that a reactor i
trip is initiated concurrently with a safety injection. The setpoint for this trip is ioentical to the safety injection setpoint.
I' Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an l
excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setpoint is sufficiently below the full load operating point so as not to interfere with normal operation, but l
still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at s 200 psis this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.
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t ARKANSAS - UNIT 2 B 2-4 Amendment No. 44,448, E ______ _ _____ _____
1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES a.*
RCS Cold Leg Temperature-Low 2 490*F b.
RCS Cold Leg Temperature-High s585'F c.
Axial shape Index-Positive Not more positive than +0.6 l
d.
Axial Shape Index-Negative Not more negative than -0.6 e.
Pressurizer Pressure-Low sl785 psia f.
Pressurizer Pressure-High s2415 psia g.
Integrated Radial Peaking Factor-Low 2 1.28 h.
Integrated Radial Peaking Factor-High 57.00 l;
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Quality Margin-Low
$0 i
Steam Generator Level - High The steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture :arry over. This trip's setpoint does not correspond to a safety I
Limit and no credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.
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l-I ARKANSAS - UNIT 2 B 2-7 Amendment No. 34,4G,-M,M,
E-3/4.2 POWER DISTRIBUTION LIMITS
- BASES I
3/4.2.1 LINEAR NEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two' core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in_the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The COLSS performs this. function by continuously monitoring-the core power l
' distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.
The.coLSS calculated core power and the COLSS calculated core power I
operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that i
the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steaily state operation. Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power
- operating limit being exceeded.
In the event this occurs, COLSS alarms i
will be annunciated.
If the event which causes the COLSS limit to be j.
exceeded results in conditions which approach the core safety limits, a
t reactor trip will be initiated by the Reactor Protective Instrumentation.
The COLSS calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by COLSS is greater than or equal to that existing in the core. To ensure that'the design margin to safety is maintained, the COLSS computer program includes uncertainties associated with planar radial peaking measurement, engineering design factors, state parameter measurement, software algorithm modeling, computer processing,' rod bow, and core power measurement.
Parameters required to maintain the operating limit power level based on linear heat rate, margin'to DNB and total core power are also monitored j
by the CPCs. Therefore, in the event that the COLSS is not being used,
. operation within the limits specified in the CORE OPERATING LIMITS REPORT can be maintained by utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPCs.
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s ARKANSAS - UNIT 2.
B 3/4 2-1 Amendment No. G4,49,Mh
POWER DISTRIBUTION LIMITS nkSES P ilt/Puntilt is the ratio of the power at a core location in the t
presence of a tilt to the power at that location with no tilt.
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.3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable udnimum DNBR throughout all anticipated operational-occurrences.
Operation of the core with a DNBR at or above.this limit-provides assurarme that an acceptable udnimum DNBR will be maintained in the event of any anticipated operational occurrence.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits.. The COLSS perforum this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR. The COLSS calculation of core power I
operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence-level that the core power at which a DNBR of less than 1.25 could occur, as calculated by COLSS, is less than or equal to that which would actually be required in the core. To ensure that the design margin to safety is maintained, the COLSS computer program includes uncertainties associated with planar radial peaking measurement, engineering design factors, state parameter measurement, software algorithm modeling, computer processing, rod bow, and core power measurement.
i Parameters required to maintain the margin to DNB and total core power are also monitored by the CPCs..Therefore, in the event that the COLSS is
-not being used, operation within the limits specified in the CORE OPERATING LIMITS REPORT can be maintained by utilizing a predetermined DNBR as a l
function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPC.
A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to accommodate the effects of rod bow.
The amount of rod bow
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in each assembly is dependent upon the average burnup experienced by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.
In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC is then determined l
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i ARKANSAS - UNIT 2 B 3/4 2-3 Amendment No. M,M,M,44,M,M4,
CONTAINMENT SYSTEMS BASES The containment cooling system and the containment spray system are redundant to each other in providing post accident cooling of the containment atmosphere. As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the containment cooling system have been appropriately adjusted.
However, the allowable out of service time requirements for the containment spray system have been maintained consistent with that =seigned other inoperable ESF equipment since the containment spray system also provides a mechanism for removing Iodine from the containment atmosphere.
The addition of a biocide to the service water system is performed l
during containment cooler surveillance to prevent buildup of Asian clams in the coolers when service water is pumped through the cooling coils. This is performed when service water temperature is between 60'F and 80*F since in this water temperature range Asian clams can spawn and produce larva which could pass through service water system strainers.
3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
The containment isolation valves have been relocated to plant procedures.
The opening of locked or sesled closed manual and deactivated automrtic containment isolation valves on an 17tenaittent basis under administrative control includes the following considesstions:
(1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing the operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will provent the release of radioactivity outside containment.
3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintein the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.
Either recombiner unit is capable of controlling the expected Lydrogen generation associated with 1) zirconium-water reactions, 2) radiolytle decomposition of water, and 3) corrosion of metal within containment. These hydrogen control systems are consistent with the recommendations cf Regulatory Guide 1.7 "Contral of Combustible Gas Concentrations in Containment Following a LOCA", March 1
1971.
I ARKANSAS - UNIT 2 B 3/4 6-4 Amendment No. M,M4,M4, i
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MARK-UP OF PROPOSED BASES CHANGES
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
]
BASES Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits (i.e., DNBR and centerline fuel melt temperature) are not exceeded during normal operation and design basis anticipated operational occurrences.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the
)
containment atmosphere.
The Reactor Coolant System components are designed to section III of the ASME Code for Nuclear Power Plant Components. The reactor vessel, i
l steam generators and pressurizer are designed to the 1968 Edition, Summer 1970 Addenda; piping to the 1971 Edition, original issue; and the valves to the 1968 Edition, Winter 1970 Addendaal.Section III of this Code permits a l
maximum transient pressure of 110% (2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the pesign criteria l
and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the l
consequences of accidents.
Operation with a trip set less conservative I
than its Trip Setpoint but within its specified Allowable value is acceptable on the basis that the differr uae between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed l
for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.25 and 21.0 kw/ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to
[
trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip setpoints.
'" Use of a later ASME Section III code is acceptable. provided the code section(s) is reconciled.
ARKANSAS - UNIT 2 B 2-2 Amendment No. 24.M.M,4M,-
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS EASES Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at s 2370.887 psia which is below the cominal lift setting (2500 psia) of the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.
Pressurizer Pressure-Low The Pressurizer Pressure-Low trip is provided to trip ti.'
reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident.
During normal operation, this trip's setpoint is set at 2 1686.3 psia.
This trip's setpoint may be manually decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at s 200 pais this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.
Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.
The setpoint for this trip is identical to the safety injection setpoint.
Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setpoint is sufficiently below the full load operating point f : preninct:ly 000 p;i: so as not to interfere l
with nornal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is naintained at s 200 psi; this setpoint 1
increases automatically as steam generator pressure increases until the trip setpoint is reached.
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ARKANSAS - UNIT 2 B 2-4 Amendment No. 49,444, l
b-
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES a.'
RCS Cold Leg Temperature-Low 2490*F b.
RCS Cold Leg Temperature-High s585'F c.
Axial Shape Index-Positive Not more positive than +0.6 d.
Axial Shape Index-Negative Not more negative than -0.6 e.
Pressurizer Pressure-Low s1785 psia f.
Pressurizer Pressure-High s2415 psia g.
Integrated Radial Peaking Factor-Low 21.28 h.
Integrated Radial Peaking Factor-High s7.004,G4 l
1.
Quality Margin-Low sO Steam Generator Level - High 1
The Steam Generator Level - High trip is provided to protect the l
turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over.
This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.
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l ARY,ANSAS - UNIT 2 B 2-7 Amendment No. G4,43,M,M,
____3
3/4.2 POWER DISTRIBUTION LIMITS aASES 3/4.2.1 LINEAR HEAT RATE The lLaitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The COLSS perforns this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.
The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady state operation. Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.
In the event this occurs, COLSS alarms will be annunciated.
If the event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.
i The COLSS calculation of the linear heat rate limit includes appropriate J
uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by COLSS is greater than or equal to that existing in the core. To ensure that the design j
margin to safety is amintained, the COLSS computer progrmn includes sneertainties associated with clanar radial oeakina measurement, enaineerina c,esian factors, state carameter measurement, software alaorithm modelina, computer orocessina, rod bow, and core power measurement.en-Eny
- ._;nt un:::::inty f::t:: f 1.052,
- :ngin:::ing un:::tcinty f::t::
ef 1.03, : T"InMr.L "O'ern :::::-- nt un:::tcinty f::t : :f 1.02 : d ppr:prict; un;; tcinty :nd p;n lty f :t;:: for ::d i x.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB and total core power are also monitored by the CPCs.
Therefore, in the event that the COLSS is not being used, operation within the limits specified in the CORE OPERATING LIMITS REPORT l
can be maintained by utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels. The above listed I
uncertainty and penalty factors are also included in the CPCs.
h l
ARKANSAS - UNIT 2 B 3/4 2-1 Amendment No. G4,M,-144, L
l POWER DISTRIBUTION LIMITS BASES I
P ilt/Puntilt is the ratio of the power at a core location in the t
presence of a tilt to the power at that location with no tilt.
3/4.2.4 DNBR MARGIN l
The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a l
conservative envelope of operating conditions consistent with the safety I
analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences.
Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of any anticipated operational occurrence.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits.
The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.
The COLSS calculation of core power operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the core power at which a DNBR of less than 1.25 could occur, as calculated by COLSS, is less than or equal to that which would actually be required in the core. To ensure that the design margin to safety is naintained, the COLSS computer program includes uncertainties associated with clanar radial peakino measurement.
encineerino desian f actors, state parameter measurement, software alcorithm modelina, computer processina, rod bow, and core power measurement.an-Fny mec urem:7t une::tcinty feeter of 1.053, en engine: ing un;;;t inty f::ter of 1.03, e-muSRe.L PCTED measurement une::tcinty facts: Of 1.02 and cpproprist une :tcinty :nd penalty f :ter for : d her.
, Parameters required to maintain the margin to DNB and total core power j
are also monitored by the CPCs.
Therefore, in the event that the COLSS is l
not being used, operation within the lindts specified in the CORE OPERATING l
LIMITS REPORT can be maintained by utilizing a predetermined DNBR as a I
function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPC.
A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to accommodate the effects of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.
In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A cingle net penalty for COLSS and CPC is then determined i
+
j ARKANSAS - UNIT 2 B 3/4 2-3 Amendment No. G4, M, M, M, M, M-7, 1
t CONTAINMENT SYSTEMS BASES f
The containment cooling system and the containment spray system are redundant to each other in providing post accident cooling of the containment atmosphere. As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the containment cooling system have been appropriately adjusted.
However, the allowable out of service time requirements for the containment spray system have been maintained consistent with that assigned other inoperable ESF equipment since the containment spray system also provides a mechanism for removing Iodine from the containment I
atmosphere.
ZheIn addition of a biocide to the service water system is performed l
during containment cooler surveillance to prevent buildup of Asian clams in the coolers when service water is pumped through the cooling coils. This is j
performed when service water temperature is between 60*F and 80*F since in this water temperature range Asian clams can spawn and produce larva which l
could pass through service water system strainers.
3/4.6.3 CONTAINMENT ISOLATION VALVES i
The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in i
l the event of a release of radioactive naterial to the containment l
atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive
)
material to the environment will be consistent with the assumptions used in the analyses for a LOCA. The containment isolation valves have been relocated I
to plant procedures."rsccdurc 2203.005r l
The opening of locked or sealed closed manual and deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations:
(1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing the operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside containment.
3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below l
its flammable limit during post-LOCA conditions.
Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water, and 3) l corrosion of metal within containment. These hydrogen control systems are l
consistent with the recommendations of Regulatory Guide 1.7 " Control of Combustible cas concentrations in Containment Following a LOCA", March 1971.
ARFANSAS - UNIT 2 B 3/4 6-4 Amendment No. M,M7,M4, j
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