Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering...
Transcript of Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering...
![Page 1: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/1.jpg)
Zoran BilanovicACR Safety Engineering
Presented to the Nuclear Regulatory CommissionWashington, DCMay 15-16, 2003
![Page 2: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/2.jpg)
Pg 2
Outline
• Disciplines involved• Physics analysis tools• System thermal-hydraulics analysis tools• Fuel analysis tools• Fission product analysis tools• Moderator analysis tools• In-core damage analysis tools• Containment analysis tools• Atmospheric Dispersion and Public Dose tools
![Page 3: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/3.jpg)
Pg 3
Disciplines Involved
• Reactor Physics• System Thermal-hydraulics• Fuel and Fuel Channel• Containment• Atmospheric Dispersion and Public Dose
![Page 4: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/4.jpg)
Pg 4
Discipline LinksReactor Physics
SystemThermal-hydraulics
Fuel Channel
Fuel
Containment
AtmosphericDispersion Public
Dose
![Page 5: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/5.jpg)
Pg 5
Computer Code Connections
- local moderatorbehavior
Reactor PhysicsRFSP - WIMS
Thermal-hydraulicsCATHENA
Fuel ChannelCATHENA
FuelELESTRES, ELOCA,….
ContainmentGOTHICSMART
AtmosphericDispersion
ADDAM PublicDose
- header boundaryconditions
- power transients
- initial core state- coolant characteristics
- thermal-hydraulic boundaryconditions
- fuel/clad temperatures- zirc/water reaction (hydrogen)- fission product inventory
distribution- fuel failure- fission product release- fission product transport and
deposition
- pressure tube strain- pressure tube sag
- high building pressure trip- ECC conditioning signal- activity release
- weather scenario- release height/location
-initial core state-power transient
In-Core DamageTUBRUPT
ModeratorMODTURC_CLAS
![Page 6: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/6.jpg)
Pg 6
Physics Analysis Tools - 1
WIMS-AECL:
• General-purpose lattice cell code• WIMS-AECL is a multi-group transport code for reactor
lattice calculations• The lattice code uses cross-sections derived from
EBDF/B-6 data library• Fuel burnup and isotope depletion
![Page 7: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/7.jpg)
Pg 7
Physics Analysis Tools - 2
WIMS-AECL:
• Inputs:− Physical dimensions of the cell− The geometrical arrangement− The composition of the cell− Temperature and density of the materials in the cell
• Outputs:− Cell-averaged parameters are tabulated as functions of fuel
burnup, fuel temperature, and coolant density− It produces 2-group cell-averaged parameters for RFSP-
CERBERUS calculations
![Page 8: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/8.jpg)
Pg 8
RFSP:• Code to generate nominal power distributions and reactor
operations, including re-fueling and burnup steps• It calculates space-time-dependent flux and power distributions
using 2-group lattice parameters as functions of: − Fuel burnup, fuel temperature, and coolant density
• The fuel cross sections are averaged over the residence (dwell) time of the fuel at each point in the core
• Obtains bundle bowers, channel power, and bundle irradiation histories
• It is coupled with the CATHENA code to use the up-to-date thermal-hydraulics data as LOCA proceeds
Physics Analysis Tools - 3
![Page 9: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/9.jpg)
Pg 9
Physics Analysis Tools - 4
RFSP• Inputs:
− Cross sections from WIMS− Core size and mesh size− Device incremental cross sections− Coolant temperature and density, and fuel temperature from
CATHENA thermal-hydraulics code• Outputs:
− Channel and bundle power distributions within the core− Thermal and fast flux distributions within the core
![Page 10: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/10.jpg)
Pg 10
Thermal-Hydraulic Analysis Tools - 1
CATHENA• Non-equilibrium two fluid system thermal-hydraulics
code• Full network defined by user in the input files• Has D2O and H2O properties• Generalized heat transfer package:
− Multiple surfaces per thermal-hydraulic node− Models heat transfer in and between fuel pins− Has built-in temperature dependent properties (has options
for user input properties)
![Page 11: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/11.jpg)
Pg 11
Thermal-Hydraulic Analysis Tools - 2
CATHENA• Inputs:
− Geometry and layout of piping− Channel power distributions from RFSP− System controls and properties− Reactor power transients from RFSP
• Outputs:− System thermal-hydraulic behavior− Boundary conditions for detailed single channel analyses− Coolant and fuel temperatures for RFSP
![Page 12: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/12.jpg)
Pg 12
SG #1 SG #2
BHH13P1
BHH13P2
BHH13P3
BHH13P4
BHH13P5
RISE1L1
RISE1U1
RISE2L1
RISE2U1
P1AD2
P1AIN
B1IN B1OUT
RISE2L2
RISE2U2
RISE1U2
B2OUT B2IN
ROH1
RIH1
RIH2
ROH2
B1TUP B1TDN
B1MDN
B1PRE
B2TUP B2TDN
B2MDN
B2PRE
S TLF1 S TLB1
ATM1
PRESSURIZER
TANKBC
TANKBK
TANKL1TANKL2
PRSTNK
PSRLN1
PSRLN2
PSRLN3
CHAN2-1
CHAN1-1 1 2 3 4 5 1 2 3 4 5 IF1-1-
OF2-1- IF2-1-
OF1-1-
S S DN2 S S DN1 S S T2 S S T1
S S M1 S S M1
S S B2 S S P2 S S B1 S S P1 RIH ROH
S S EP1
S DRT1
S S EP2
S DRT2
5 4 3 2 1 5 4 3 2 1
SSFEEDBC
SSFEEDBC
TURB TURB
ATM3
S TLF2 S TLB2
ATM3
ATM1
P1ASUCT
P1BD2
P1ADIS
P1BIN
P1BSUCT
P1BDIS P1BD1
P1AD1
P2BD2
P2ADIS
P2BIN
P2ASUCT
P2BDIS
P2BSUCT
P2AIN
P2AD2
RIH = Inlet Header ROH = Outer Header IF = Inlet Feeder OF = Outlet Feeder PRSTNK = Pressurizer CHAN = Fuel Channel BOUT = Boiler Outlet Plenum
FEEDBC = Feed BLEEDBC = Bleed TURB = Turbine SSEP = Separator SSDN = Downcomer SSFEEDBC = Feedwater BIN = Boiler Inlet Plenum
RISE1L2
P2AD1
P2BD1
B1MUP
B1BUP
B22MUP
B2BUP
HH13P1
HH13P2
HH13P3
HH13P4
HH13P5
VALSUR3
B1KHH1-3
S TLB3 S TLF3
ATM1 ATM1
TURB TURB
S TLF4 S TLB4
ATM3 ATM3
ILT1-1 IEFB1-1 ISP1-1 OSP1-1
OEFB1-1
OLT1-1
OLT2-1 OEFB2-1 OSP2-1 ILT2-1IEFB2-1ISP2-1
CATHENA Nodalization diagram of the ACR RCS circuit
Thermal-Hydraulic Analysis Tools - 3
![Page 13: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/13.jpg)
Pg 13
Thermal-Hydraulic Analysis Tools - 4
CATHENA Single Channel Analysis• Single channel model calculates the detailed fuel and
fuel channel thermal-hydraulic transient
![Page 14: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/14.jpg)
Pg 14
Thermal-Hydraulic Analysis Tools - 5CATHENA Single Channel Analysis • Inputs:
− fuel and fuel channel geometry− fuel and fuel channel axial and radial power distribution− transient header boundary conditions from CATHENA full
circuit calculations− power transient from RFSP-CERBERUS calculations
• Outputs: − fuel boundary conditions (pressure, temperature, and fuel-to-
coolant heat transfer coefficient) for ELOCA− pressure tube temperatures and strain− initial conditions for TUBRUPT in-core damage calculations− hydrogen generation
![Page 15: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/15.jpg)
Pg 15
RIH1
IF1-1-1
IF1-1-2
IF1-1-3
IF1-1-4
IF1-1-5
ILT1-1 ISP1-1
CHAN1-1
MODERATRMODIN MODOUT
ROH2
OF1-1-5
OF1-1-4
OF1-1-3
OF1-1-2
OF1-1-1
OEFB1-1
OSP1-1 OLT1-1
IEFB1-1
Thermal-Hydraulic Analysis Tools - 6
CATHENA Nodalization diagram for a single channel
![Page 16: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/16.jpg)
Pg 16
CATHENA Nodalization diagram of the ACR channel
Thermal-Hydraulic Analysis Tools - 7
![Page 17: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/17.jpg)
Pg 17
Fuel Analysis Tools - 1
ORIGEN• Calculates the time-dependent concentrations and
inventories of a large number of isotopes• The code takes into account the simultaneous
generation and depletion of isotopes from the generation and depletion through neutron transmutation, fission, radioactive decay, input feed rates, or physical or chemical removal rates
![Page 18: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/18.jpg)
Pg 18
Fuel Analysis Tools - 2
ORIGEN• Inputs:
− Cross sections from WIMS− Fuel composition− Power / burnup history from RFSP
• Outputs: − Initial total inventories of fission products
![Page 19: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/19.jpg)
Pg 19
Fuel Analysis Tools - 3
ELESTRES • Models the thermal and mechanical behavior of an
individual fuel element during its irradiation life under normal operating conditions
• Composed of two models:− one-dimensional fuel performance model:
• thermal model for temperature calculations• microstructural model for fission gas and associated
calculations− two-dimensional stress analysis model:
• used to calculate axisymmetric deformations of the fuel pellet• Supplies the initial fuel conditions
![Page 20: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/20.jpg)
Pg 20
Fuel Analysis Tools - 4
ELESTRES • Inputs:
− Fuel geometry and composition, and− Power / burnup history from RFSP
• Outputs: − Heat generation rates within the fuel pellets for input to ELOCA− Clad strain for input to ELOCA− Fuel-to-clad gap conductance for input to ELOCA− Distribution of fission products (grain bound, grain boundary,
and fuel-to-clad gap) for input to SOURCE
![Page 21: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/21.jpg)
Pg 21
Fuel Analysis Tools - 5
ELOCA• A transient fuel behavior analysis code for calculating
the fuel element thermo-mechanical behavior during a postulated transient
![Page 22: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/22.jpg)
Pg 22
Fuel Analysis Tools - 6
ELOCA • Inputs:
− heat generation rates, clad strain, and fuel-to-clad gap conductance from ELESTRES
− fuel boundary conditions (pressure, temperature, and fuel-to-coolant heat transfer coefficient) from CATHENA
− power transient from RFSP-CERBERUS calculations• Outputs:
− fuel and clad temperatures for input to SOURCE− fuel failure time− the extent of clad oxidation for input to SOURCE
![Page 23: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/23.jpg)
Pg 23
Fission Product Analysis Tools - 1
SOURCE• Calculates the amount and type of fission product
released from fuel• Some of the phenomena modeled:
− Diffusion, grain boundary sweeping, vapor transport, gap transport, fuel/zircaloy interaction, temperature transients, grain boundary separation, etc.
![Page 24: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/24.jpg)
Pg 24
Fission Product Analysis Tools - 2
SOURCE• Inputs:
− fuel temperature transients from ELOCA− fuel boundary conditions (pressure, temperature, and fuel-to-
coolant heat transfer coefficient) from CATHENA− power transient from RFSP-CERBERUS calculations
• Outputs: − fission product release transient – amounts and species
![Page 25: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/25.jpg)
Pg 25
Fission Product Analysis Tools - 3
SOPHAEROS• Calculates the fission product deposition and transport
through the heat transport system components and piping
• The code employs a pseudo-kinetic thermochemical-equilibrium numerical solution scheme using deviation from the equilibrium state as driving force to calculate fission-product specification and volatility
![Page 26: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/26.jpg)
Pg 26
Fission Product Analysis Tools - 4
SOPHAEROS• Inputs:
− fission product release transients from SOURCE− RCS boundary conditions (steam pressure and temperature,
materials and temperatures) from CATHENA• Outputs:
− fission product retention within the RCS, and deposition on the RCS materials
− the amount and species of the fission products released into containment
![Page 27: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/27.jpg)
Pg 27
Moderator Analysis Tools - 1
MODTURC-CLAS• General purpose CFD code used to assesses the
moderator velocity and temperature distribution• Consists of coupling ACR moderator related specific
modules with the general purpose CFD code• Calculation of poison concentration distributions after
Shut Down System 2 (SDS2) initiation
![Page 28: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/28.jpg)
Pg 28
Moderator Analysis Tools - 2
MODTURC-CLAS• Inputs:
− Moderator circuit geometry and conditions− Heat load to the moderator during normal operation and
accidents conditions obtained from CATHENA• Outputs:
− Moderator subcooling margin− Distribution of moderator poison after SDS2 initiation
![Page 29: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/29.jpg)
Pg 29
In-Core Damage Analysis Tools - 1
TUBRUPT• A code used to determine the pressure transients
within the calandria vessel caused by an in-core LOCA (channel rupture)
• Assesses the damage to in-core components as a result of an in-core LOCA
![Page 30: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/30.jpg)
Pg 30
In-Core Damage Analysis Tools - 2
TUBRUPT• Inputs:
− fuel channel conditions at the time of channel failure (pressure and temperature) from CATHENA
− core and channel geometry• Outputs:
− assessment of in-core damage (calandria vessel integrity, adjacent channel integrity, and shut-off rod guide tube integrity)
![Page 31: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/31.jpg)
Pg 31
Containment Analysis Tools - 1
GOTHIC• 3-D 2-fluid thermal-hydraulic code to determine the
transient conditions inside the reactor building• Track hydrogen as a non-condensable gas within
containment• Lumped parameter model for peak pressure
assessment• Lumped parameter model in combination with 3-D
model (in the vicinity of the discharge) for hydrogen assessment
![Page 32: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/32.jpg)
Pg 32
Containment Analysis Tools - 2
GOTHIC• Inputs:
− break discharge and hydrogen release transients from CATHENA
− containment geometry and heat sinks• Outputs:
− Thermal-hydraulic conditions to SMART− Peak containment pressure− Hydrogen distribution within containment
![Page 33: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/33.jpg)
Pg 33
Containment Analysis Tools - 3
SMART• Calculates the radionuclide behavior inside the reactor
building and calculates the radionuclide release to the outside atmosphere via various release paths
• An aerosol general dynamics equation is solved to calculate aerosol size distribution as a function of space and time
• Calculates the speciation of iodine within containment
![Page 34: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/34.jpg)
Pg 34
Containment Analysis Tools - 4
SMART• Inputs:
− Thermal-hydraulic conditions from GOTHIC− Fission product releases from SOPHAEROS
• Outputs: − Fission product releases to atmosphere to be used in dose
assessment by ADDAM
![Page 35: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/35.jpg)
Pg 35
Dose Assessment Tools - 1
ADDAM• Uses Gaussian dispersion model to calculate
dispersion factors• Applies corrections due to nearby buildings, release
height, and deposition
![Page 36: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/36.jpg)
Pg 36
Dose Assessment Tools - 2
• ADDAM:• Inputs:
− Fission product releases from SMART− Thermal-hydraulic conditions from GOTHIC
• Outputs: − Internal dose due to inhalation and skin absorption of
radioactive material from the cloud− External dose from exposure to radioactive material in the
cloud (cloud shine)− External dose due to exposure to radioactive material from
ground deposition (ground shine)
![Page 37: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/37.jpg)
Pg 37
Summary
• ACR-700 safety analysis methodology is based on computer codes that have been developed and used for several decades in CANDU safety analysis at AECL and across the Canadian industry
• The AECL computer codes cover the full scope of the ACR safety analysis needs
![Page 38: Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering Presented to the Nuclear Regulatory Commission Washington, DC May 15-16, 2003 Pg 2](https://reader035.fdocuments.in/reader035/viewer/2022081619/60ebd7344c2287288825c543/html5/thumbnails/38.jpg)
Pg 38