Project DENIDIA: A MCNP code as a tool for modelling
and simulation of gamma-ray tomography
Computer Engineering Department Technical University of Lodz18/22 Stefanowskiego, 90-924 Lodz, Poland [email protected]
Volodymyr Mosorovoutgoing researcher at University of Bergen,
Norway
February, 2009
Outline
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1.Physical background of Radiation measurement
1.Radioisotope gauges: modeling and simulation
1.MCNP model of the UIB 85 channel gamma ray tomograph
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The Lambert-Beers Law is the simplification of real world,
In practice a particle transport has a stochastic nature,
One of the powerful method for simulations of particle transport is Monte Carlo simulation
This is a topic of todays present class.....
Reminding
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Stochastic nature of photons transport
The main reason:
Different mechanisms of Gamma-Ray Interactions with matter
(see a previous class)
Deterministic methods solve the transport equation for the average particle behavior.
Monte Carlo method obtains answers by simulating individual particles. The average behaviour of particles is then inferred from the average behaviour of the simulated particles.
Monte Carlo Method vs. Deterministic Method (e.g. Lambert-Beers Law)
The Monte Carlo Method
Monte Carlo is used to simulate statistical processes theoretically (like the interaction of nuclear particles with materials)and is particularly useful for complex problems that cannot be modeled by computer codes that use deterministic methods.
The individual probabilistic events that comprise a process are simulated sequentially.
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MCNP (Los Alamos National Laboratory)
EGS4 (http://slac.stanford.edu/egs)
)
Geant (http://geant4.web.cern.ch/geant4/)
)
Available software packages for simulation radioactive particle transport:
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MCNP What is the MCNP?
MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport.
What MCNP can doradiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety,accelerator target design, fission and fusion reactor designetc.
From MCNP workshop
MCNP details:
Version: 4A ~1994 4B ~1997 4C1 ~2000 4C2 ~ fixed bugs 4C3 ~ minor fix MCNPX 5 ~ rewritten to include VisEd (400 person years
of development)
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MCNP5 package
Distributed through RSICC (Radiation Safety Information Computational Center) at Oak Ridge National Laboratory, USA,
RSICC's mission is collection of computer code technologies regarding radiation transport and safety to meet the needs of the international community.
Code Contribution is not free
Current Director: Bernadette L. Kirk
MCNP5 package cd
Computer Platforms:UNIX: Linux, HP, SGI, SUN, IBM
Windows/DOS
Written in FORTRAN 90
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MCNP5 packageAdvantages: It is more accurate model of particle transport
Disadvantages: very computationally extensive, poor interface, time-consuming
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MCNP code conceptPlace initial photons parameters on stack
Pick up energy, position, direction, geometryof current particle from top of the stack
Photon energy< cutoff?
Determine distance to next interactionTransport photon taking into acount geometry
Hasphoton left volume
of interest?
Determine type of interaction; Photoelectric,Compton, Rayleigh, Pair production
Determine energies and directions of resultantparticles and store parameters on stack
Terminate history
N
Y
Y
N
Is stack empty?
Y
N
MCNP details:
MCNP is a 3-D Monte Carlo code capable of tracking 34 particle type (including 4 light ions)
Energy range: Neutron: 0.01 MeV 20 MeV Photon: 1 keV 100 GeV Electron: 1 keV 1 GeV
Related Codes:
VisEd: 2D graphics are included
Moritz: 3D does not include MCNP, takes files from MCNP, visualizes geometry and particle tracks
MCODE: Couples MCNP to Origen for burnup
Monteburns: Couples MCNP to Origen for burnup
NJOY: Generates MCNP libraries
MCNP Basic Input
The basuc units used in MCNP are: Lengths in centimeters Energies in MeV Times in sec Temperatures in kT Atomic densities in units of atoms/barn-cm Mass densities in g/cm^3 Cross section in barns Heat number in MeV/collision Atomic weight ratio based on a neutron mass of
1.00866497 amu
MCNP Basic Input Input file:
Cell card Surface card Data card
Prototype
MCNP Cells
Cells are defined as volumes of space bounded by surfaces
All of space must be defined At least one cell will describe the problem Repeated structure and lattice ability
Cell commands
Cells are basic geometry unit Cartesian coordinate system Surfaces are used to define space by combing
Sign defines surface sense (+ or -) Three available operators
Interaction space Union : Complement #
Surface command
Example
J t params
Surfa
ce n
umbe
rm
nem
onic
dimen
sions
Surf # Name Parameters
1 cz 20 $ infinite z cylinder with radius 20 cm
Surf # Name Parameters
1 so 11 $ sphere with radius 11 cm
$ equation: x2+y2+z2-r2=0
Surface sense
F(x,y,z)=SWhere F is the function of surfacex,y,z are the coordinates of and arbitrary pointS sense of the point: positive zero (point is on the surface) negative
Surface sense cd
point 1
0+0+0-9 = -0 : negative sense
point 20+36+0-9 = 27 : positive sense
Cell 1Cell # Mat # Den Surface
1 3 1.65e-2 -1
Cell 2Cell # Mat # Den Surface
2 3 1.65e-2 +1
Y
Sphere R=3
(0,0,0)
Point 1
Point 2 (0,-6,0)
Cell 1
Cell 2
Boolean Operator
Three available operators Interaction space Union - : Complement - #
Intersection
Logical AND Only space between
BOTH criteria are true
Cell # Mat # Den Surface1 4 XXXX -1 +22 5 XXXX +1 -23 6 XXXX -1 -2
Surface 1
Surface 2
IIII II
Intersection: example
Cell ISurface 1 Surface 2
Surface 3
Surface 4
x
y
Cell I
Surface Sense
Positive Negative
1 X
2 X
3 X
4 X
Boolean Union
Logical ORCell # Mat # Den Surface
1 4 XXXX -1 : -2
Surface 1
Surface 2
Cell I
Boolean Union: Example
Cell ISurface 1 Surface 2
Surface 3
Surface 4
x
y
Cell II
Surface Sense
Positive Negative
1 X
2 X
3 X
4 X
Cell II
Complement Operator
# n means that the description of the current cell is the complement of the description of cell n
Cell card1 +1 -2 +3 -42 -1 : +2 : -3 : +4Cell 2 also can be described as2 # 1
Cell ISurface 1 Surface 2
Surface 3
Surface 4
x
y Cell II
Summary of MCNP Cells
Cells are basic geometry unit Cartesian coordinate system Surfaces are used to define space
Sign defines surface sense (+ or -) Boolean interaction, union, complement
parentheses control order of operations At least one outside world cell
Data Cards
Class Mnemonic
Materials mImportance impSource sdef Tallies fProblem cutoffs nps
Any card other than cells and surfaces goes to data cards
Materials card
Mn ZAID1 fraction1 ZAID2 fraction2 ZAID = element or nuclide identifier : ZZZAAA
ZZZ = atomic numberAAA = atomic mass number
Example: 92235 (235U) 8016 (16O ) 29000 (natural Cu)Fraction: positive = atomic fraction ZAIDi
negative = weight fraction of ZAIDi
Example: M1 07014 -0.767 08016 -0.233
All the available nuclei in MCNP can be found in the file: xsdir
Importance card
Each cell must have an importance, used for variance reduction
Importance of 0 will terminates particles, outside world cell usually is 0
It can be in data card block, Example
2 -0.132 -7:8:-9 imp:n=1 3 -1.235 1 2 3 imp:n=0
Source card
General sdef Surface User-supplied:
Energy Time Direction u v w Position x y z Particle type Weight
SDEF Card
Example: sdef ERG=D1 POS=x y z CEL=m RAD=D2 EXT=D3 AXS=I j kSI1 H 0.001 0.01 0.1 1 10 SP1 10 0.1 0.01 1 15SI2 r1
SI3 a
Cell mx,y,z
i,j,kr1
a
0. 0010. 010. 1
110
100
0. 001 0. 01 0. 1 1 10Energy(MeV)
Prob
abil
ity
Tally Card (detector)
MCNP execution
mcnp i=inp01 o=out01 [options]File name DescriptionINP input file nameOUTP ASCII output file name
Option operation DefaultI process input file *P Plot geometryX process x-section *R particle transport *Z plot tally results
plot cross section
MCNP Execution
Examples;mcnp inp=test1 outp=test1o ipmcnp i=test1 ixz
