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