Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering...

Zoran BilanovicACR Safety Engineering

Presented to the Nuclear Regulatory CommissionWashington, DCMay 15-16, 2003

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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

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Disciplines Involved

• Reactor Physics• System Thermal-hydraulics• Fuel and Fuel Channel• Containment• Atmospheric Dispersion and Public Dose

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Discipline LinksReactor Physics

SystemThermal-hydraulics

Fuel Channel

Fuel

Containment

AtmosphericDispersion Public

Dose

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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

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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

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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

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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


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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

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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)

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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

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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


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CATHENA Single Channel Analysis• Single channel model calculates the detailed fuel and

fuel channel thermal-hydraulic transient

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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

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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


CATHENA Nodalization diagram for a single channel

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CATHENA Nodalization diagram of the ACR channel


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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

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ORIGEN• Inputs:

− Cross sections from WIMS− Fuel composition− Power / burnup history from RFSP

• Outputs: − Initial total inventories of fission products

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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

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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

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ELOCA• A transient fuel behavior analysis code for calculating

the fuel element thermo-mechanical behavior during a postulated transient

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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

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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.

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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

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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

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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

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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

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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

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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

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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)

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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

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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

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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

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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

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Dose Assessment Tools - 1

ADDAM• Uses Gaussian dispersion model to calculate

dispersion factors• Applies corrections due to nearby buildings, release

height, and deposition

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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)

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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

Zoran Bilanovic ACR Safety Engineering Presented to the ...Zoran Bilanovic ACR Safety Engineering...

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