MCNP code

Siriyaporn Sangaroon 25 September 2014

Introduction Simulation using MCNP Visual Editor Variance Reduction

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Introduction

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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 can not be modeled by computer codes that use deterministic methods.

The individual probabilistic events that comprise a process are simulated sequentially.

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Number of publications on various MC codes

Modern Radiation Transport Codes • GEANT (CERN) • MCNP (LANL - USA) • SCALE / Morse / KENO (ORNL – USA) • TRIPOLI (France) • Answers / Monk / McBend (UK) • PHITS (Japan) • MCU (Russia) • SHIELD (Russia)

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The Monte Carlo Particle Transportation Grand Prix, during last 10 years

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Used in a large number of different research fields, including applied nuclear physics, nuclear physics, elementary particle physics, medicine and space physics. The simulation and modelling tools: PENELOPE - Monte Carlo simulation package for photon and

electron transport (www.nea.fr) MCNP - Monte Carlo package for neutron and photon simulation

(www.lanl.gov) GEANT - Simulation package for particle transport trough matter

(geant4.cern.ch) FLUKA - Calculation of particle transport and interactions with

matter (www.fluka.org)

Tools for modelling and simulation of particle transport

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PENELOPE

A Code System for Monte Carlo Simulation of Electron and Photon Transport written by Francesc Salvat, Jose M. Fernandez-Varea, Joseph Sempau from ECM University Barcelona (2001). Distributed through the NEA data bank

Fortran based simulation code → possible to link with cernlib (doesn't require knowledge of fortran for simulation only)

Used mainly in medical physics Easy to install and operate. Doesn't require large resources.

(~250MB)

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GEANT

Geant4 is a software using an Object-Oriented environment (C++) Many requirements taken into account, from heavy ion physics to

medical applications A large degree of flexibility is provided Toolkit

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FLUKA

1962: MC code(s) for high-energy proton beams J. Ranft (Leipzig) and H. Geibel (CERN)

1970: Study of event-by-event fluctuations in calorimeters => FLUktuierende Kaskade, Mainly used for radiation shielding studies

1970-1987: Development by J. Ranft and J.H. M๖hring (Leipzig) with significant contributions from P. Aarnio and J. Routti (Helsinki), J.M. Zazula (Cracow) and A. Fass๒ and G.R. Stephenson (CERN)

1989-: A. Ferrari and P.R. Sala (INFN Milano), together with A. Fass and J. Ranft, transforms FLUKA into a general purpose MC code

2003: CERN-INFN Collaboration Agreement 2006: Many improvements, free format input, nice tools… 2011: Gfortran option available

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MCNP

Developed at Los Alamos over 5 decades 100s of man years development Stands for Monte Carlo Neutrons and Particles 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

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My work on MCNP Simulation (2013-2014)

Liquid scintillator (together NRESP code): Response function, Efficiency

Neutron camera for MAST

Neutron 2.45 MeV (and g-rays)

Conceptual design of the Neutron Camera Upgrade for MAST Upgrade

Experimental hall, MAST

NCU Geometries Neutron flux, emissivity

profile Shielding (materials) Scattered neutron

(in/back scattered) Background g-ray

(neutron capture) … 11

Simulation using MCNP

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MCNP Simulation

Last release of MCNP5 (version 1.60) (NPE) Last release of MCNPX (version 2.7.0)

o Capable of tracking 34 particle types o Energy range:

Neutron: 0.01 MeV – 20 MeV Photon: 1 keV – 100 GeV Electron: 1 keV – 1 GeV

MCNP6 o Essential features of MCNPX and MCNP5 available in MCNP6 o MCNP6 Version 1.0 released Aug 2013.

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Applications

Homeland Security Detector Responses, Including

Electrons Medical Shielding -- neutron and photon Reactor Physics Neutronics Well Logging -- source, detector

Health Physics Criticality Safety Magnetic Fusion Neutronics Activation and Decommissioning Space and Accelerator Energy Deposition

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Documentation

Compiling

Executables available for all officially supported systems –Unix (Sun) –Linux (Intel, PGI) –Windows XP, Vista, Windows 7 (Intel)

How to install (Windows)

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- After installed, copy folder “MCNP_DATA” to C:\MCNP\

Execution

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Execution

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Input file

1. Word pad 2. The Visual Editor for Monte Carlo N-Particle : code for visually creating

and graphically displaying input files for MCNP

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Input file

TITLE CARD

CELL CARDS

SURFACE CARDS

DATA CARDS

Whatever isn’t a surface or cell card. Source sdef, kcode Tallies f2, f4, f6, … Materials m1, m2, … Variance Reduction imp, wwg, … Problem terminate nps, ctme Peripheral cards mode, phys, …

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Surface Cards

*MCNP5_manual_VOL_II page 3-13 21

Surface Card Format

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Cell Cards

Once you have defined surfaces, then you combine those surfaces into cells using the intersection and/or the union of surfaces.

Cells are the basic unit of MCNP geometry Cells are defined by Surfaces Cartesian Coordinate System Must account for all phase space Every xyz point will lie either on a surface or within a uniquely

defined cell. At least one cell will describe the “outside world”, exterior to

the problem cells (with importance of zero).

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All phase space defined

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Cell Card Format

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Boolean Intersection

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Boolean Union

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Order Of Operations

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Example (2 cubes nested)

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Macrobodies Simplify Cell Descriptions

With Macrobodies a cube is a single surface Inside a macrobody is negative sense Outside a macrobody is positive sense

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Macrobodies

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2 Cube Example with Macrobodies

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MACROBODIES (cont)

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Macrobody Limitations

Limited Number of Macrobodies May want / need to use both surfaces & bodies Still need to understand Boolean Operators Macrobodies have eccentricities

o Specifying a facet for SSR & SSW o Specifying a facet for a flagged surface (fatal) o Items that may involve a facet in PTRAC o Surface sense changed for some Macrobody facets

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Data Cards

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Mn (Material) Card

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Importances

Either an IMP or WWN card is required; most of the other cards are for optional variance reduction techniques

Surface

Surface wizard Macro body

The Visual Editor for Monte Carlo N-Particle

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*Note: Visual Editor

Cell

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*Note: Visual Editor

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More Problem Cutoff Options STOP card/stop<inp> file

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MODE Card

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MCNP Particles

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Auxiliary Input Files (READ Card)

• SSR/SSW

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Source Description

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SDEF Description

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SDEF1 Input File

Plot source

Plot track

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Checking The Source (output)

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First 50 Particles

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SDEF Description

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SI, SP, SB, and DS CARDS

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Specific SDEF Example on Plasma Fusion

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Tallies

Definitions

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Tally Types

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Basic Tally Format

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Additional Tally Capabilities

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Visual Editor The Visual Editor for Monte Carlo N-Particle : code for visually creating and graphically displaying input files for MCNP

* The default location is C:\MCNP

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What is the Visual Editor?

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Complete Interface for MCNP

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Complete Interface for MCNP

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Example geometry

BOX (x=0.525 cm, y=0.03 cm, z=12cm) Electron source

BOX (x=0.525 cm, y=0.03 cm, z=0.1cm) Attenuator (Tungsten)

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Source (sdef)

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Plot Tracks

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Tally

Unit 1/cm2 per Electron flux (5x10^8)

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Tally

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Variance Reduction

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Definitions

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Ten Statistical Checks (Output file)

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Variance reduction techniques used to improve efficiency

Either an IMP or WWN card is required; most of the other cards are for optional variance reduction techniques

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Cell Importance Cards: IMP

The importance of a cell is used to terminate the particle’s history if the importance is zero, for geometry splitting and Russian roulette to help particles move to more important regions of the geometry.

Simple Geometry

Slab of lead divided into 10 cell by planes

Y axis

Source: sdef sur 10 vec 0 1 0 dir=d1 erg 100 par p

Tally: f1:p 20 (Current integrated over a surface, unit in particles)

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Example 1: imp:p 1 10r 0 All cell in side the universe have the important = 1

FOM = 290

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Example 2: Random the cell importance value

Source

Tally 21 1 -11 -30 10 -11 imp:p=2 22 1 -11 -30 11 -12 imp:p=4 23 1 -11 -30 12 -13 imp:p=8 24 1 -11 -30 13 -14 imp:p=16 25 1 -11 -30 14 -15 imp:p=32 26 1 -11 -30 15 -16 imp:p=64 27 1 -11 -30 16 -17 imp:p=128 28 1 -11 -30 17 -18 imp:p=256 29 1 -11 -30 18 -19 imp:p=512 30 1 -11 -30 19 -20 imp:p=1024 31 0 (-10:30:20) -31 imp:p=1 32 0 31 imp:p=0

Cell 21 ........... 30

FOM = 4806 (very high) but it did not pass 1 of 10 statistical checks

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Weight windows: Cell-based (from previous example) Using Weight Window Generation: WWG

wwg 1 21

problem tally number

invokes cell- or mesh-based weight window generator (typically a source cell)

*.e file are generated to use for the next run

wwe:p 1.0000E+02 wwn1:p 5.0000E-01 1.6259E-01 4.5190E-02 1.3010E-02 4.0550E-03 1.3200E-03 5.2000E-04 2.1000E-04 1.0500E-04 5.0000E-05 0.0000E+00 -1.0000E+00

From the first run

Run Statitic FOM

1 7/10 273

2 10/10 4091

3 10/10 4674

Use *.e file from run #1

Use *.e file from run #2

Weight Window Cards Weight windows can be either cell-based or mesh-based.

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Weight windows: Mesh-based

Geometry Mesh for weight window

AIR

HDPE

LEAD

2.5 MeV neutron from outer surface of the sphere

Neutron capture in HDPE

cell flux (F4)

Weight windows: Mesh-based

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weight window

Neutron: 7th run

Gamma: 7th run

Neutron: 1st run

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neutron gamma

With weight window

ctme 1 (min)

No weight window (imp = 0, 1)

neutron gamma

Tally plots

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Note

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Beam* SDEF Example