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