simulation using mcnp - uppsala university
TRANSCRIPT
Simulation using MCNP
Sean Conroy
Why do a simulation?
Computers are cheap, detectors are expensive
Can optimise an experiment
Can generate response functions for an instrument
Can be pretty educational
MCNP
Developed at Los Alamos over 5 decades
100s of man years development
Stands for Monte Carlo Neutrons and Particles
Can track its lineage back to the very first Monte Carlo code (Ulam + von Neumann 1947)
With the associated cross section data, this represents our understanding of radiation interaction
Basic method
Generate particles with arbitrary energy, direction and species
These are tracked through arbitrary 3D geometry
Physics of the interactions of the particles well modelled
Use tallies to see what went where
Running MCNP
Create an input file
File contains:
• Geometry of the world in the problem
• Source definition
• Tallies to monitor results
• Some admin (how long to run, etc)
Feed the file to MCNP, await output.
Most models are simple
Transport flask for strong sources
Mostly polythene, wrapped in steel with a lead core
Californium neutrons released at centre of the flask
Radiation dose around the flask wanted
Remove cork from flask
Models can be pretty complicated
Californium source at JET
Applications
Spallation targets, nuclear waste transmutation, accelerator driven power, neutron, gamma or proton radiography, shield design, medicalphysics, single event upsets, reactors,safeguards, criticality, nuclear material detection, experiment design, dosimetry…..
The list goes on and on
General comments
Units: cm, shakes (10-8 s) and MeV (1.6x10-13 J)
MCNP hates tabs
Only 2 blank lines allowed, dividing sections
MCNP input files have up to 80 characters per line
The geometry must be complete, all space must be described somewhere
How to do MCNP
• Describe the surfaces in the problem
• Make cells from surfaces
• Fill the cells with materials
• Specify the source of particles
• Specify what to measure
• Specify number of particles to launch
• Run the job
Sample input file has 3 sections
C Backscatter model10 0 10 IMP:N=020 0 -10 30 40 IMP:N=130 0 -40 IMP:N=140 0 -20 40 IMP:N=150 1 -7.8 -30 20 40 IMP:N=1
10 S 0 0 0 100020 S 0 0 0 3030 S 0 0 0 9040 RCC 0 0 25 0 0 70 2
MODE NM1 26000 -1SDEF ERG=2.500001 POS=0 0 0 RAD=20F5:N 0 0 150 5E0 0.1 24I 2.6CUT:N J 0.01NPS 100000
Cells section
Surfaces section
Materials, sources and tallies section
Surfaces
MCNP uses simple surfaces, planes, spheres, cones, torii, cylinders
Every surface has a positive and negative sideFor spheres and cylinders, inside is negative and
outside is positiveA sphere centred on (2,0,0) of radius 1cm:
20 S 2 0 0 1A cylinder of radius 1cm, 2cm thick at (-1,-1,-1)
30 RCC -1 -1 -1 0 0 2 120 and 30 are the surface numbers
Last two lines look like
XZ plane XY plane
Cells
Cells are volumes defined by combinations of surfaces and the material in them and their importance
Only one type of material per cellA typical vacuum cell (Material 0 is vacuum):Material Importance
20 0 -20 IMP:N=1
Cell number Surface used
Cells
If a cell has material in it then the line looks like
30 1 -2.2 -30 IMP:N=1
The -2.2 means the density of material type 1 in the cell is 2.2 g cm-3
The rest of the universe
MCNP needs a boundary around the problem geometry beyond which the importance is zero
A simple geometry would be:
10 0 10 IMP:N=0 $ Anything beyond a $
20 0 -10 20 IMP:N=1 $ is a comment
30 0 -20 IMP:N=1
10 S 0 0 0 1000 $ Edge of the universe
20 S 1 1 1 10 $ A ball in the model
Materials, sources, tallies
MCNP can run neutrons, photons or both
Determined by the MODE line
MODE N $ Neutrons
MODE P $ Photons
MODE N P $ Neutrons and Photons
materials
Build up materials out of available libraries
M1 26000 -1 $ Pure natural iron
M2 1001 -0.548 $NE213 scintillator
6000 -0.452
M3 32000 -1 $Pure Germanium
Mn ZZZAAA Fraction of nuclei
Sources
By default, a point source at (0,0,0) isotropically radiating 14 MeV neutrons
SDEF line allows different sources to be made
SDEF ERG=2.5 $ 2.5 MeV neutrons
SDEF POS=1 1 0 $ 14 MeV neutrons at (1,1,0)
SDEF VEC=0 0 1 DIR=1 $ Monodirectional
Tallies and energy bins
Tallies tell you what went where
F4:N 30 $Flux given by volume average tally
F8:P 30 $ Energy deposited per event
Both apply to events in cell 30
F8 tallies only work well with photons
E0 0.1 25I 2.6 $ 0.1 MeV bins up to 2.6 MeV
E8 0 1e-5 1e-3 1999I 2.001 $ F8 tallies should be like this
Cuts and how many particles
CUT:N J 0.1 $ Kill neutrons below 0.1 MeV
NPS 10000 $ Run 10000 particles
Alternatively
CTME 0.1 $ Run MCNP for 0.1 minutes
How to run MCNP
Make your input file on a computer
Make sure your first name is in the first line
In a browser open
http://lillekis.tsl.uu.se:8766/cgi-bin/mcnp.cgi
Use the browse button find your input file
Press submit
MCNP will run then return the answer
MCNP output
On the handout, several sections exist.
Copy of input file
Some tables, eg, masses + cross sections
Particle history summaries
Tally results with energy bins and errors
Overall tally summary
What to simulate?
High Purity Germanium detector (as PENELOPE)
7.5 cm diameter, 7.5 cm height cylinder
0.511 & 1.275 MeV photons fired into the axis from a point 5 cm above the crystal
Make a pulse height spectrum for 106 photons
Plot spectrum, describe features present
What to simulate II
Using the basic model, investigate the following:
Efficiency as function of Photon energy (up to 10 MeV) and what happens to the escape peaks?
Isotropic source instead of monodirectional?
What happens when the beam goes off axis?
What happens when different sized detectors are used?
What to simulate III
Optional
Try adding the extra material around the detector as in the PENELOPE model. What is the effect on the spectrum of the isotropicsource?
Report
Write an individual report on the simulations.
4-6 sides should be enough
Do include the input file for the basic model