MCNPX Wilson

An introduction to MCNP/MCNPX 

Adelaide 

20 June 2011 

Dr. Puthenparampil Wilson 

Senior Medical Physicist (Royal Adelaide Hospital, Adelaide) 

Senior Lecturer (University of South Australia, Adelaide) 

Associate Fellow (PTCRi, Oxford) 

1

A. Overview of MCNPX 

B. Input deck format 

Contents 

C. Installation and running the code 

D. Applications in particle therapy 

2

Overview and introduction 

3

MCNPX- Introduction 

� A General Monte Carlo N- Particle transport code 

� Capable of tracking photons, electrons , light ions and heavy ions 

� 3D tracking 

� Wide range of energies ( keV- GeV ) 

� Written in FORTRAN 90 

� Platforms: Unix, Linux, WINDOWS, OS X 

4


� Capable of tracking 35 particle type 

� Including 4 light ions and heavy ions (He, B,C O etc..) 

� Wide energy range 

� Photons (keV- GeV) 

� Electron (keV- GeV) 

� Protons (MeV- GeV) 

5

� Applications of MCNPX 

� Medical physics 

� Particle therapy 

� Nuclear Medicine 


� Radiation protection and dosimetry 

� Nuclear criticality studies 

� Accelerator target design 

� Fission and Fusion reactor design 

� Radiation shielding 

6

MCNPX- History 

� Started the development at Los Alamos National Laboratory more 

than 60 years ago 

� 1991- MCNP 4 

� 2003 - MCNP 5 

� 2011- MCNPX 2.7.0 (Monte Carlo N-Particle Transport Code eXtended) 

� MCNPX 2.7.0 is the current public release 

� MCNP6 beta version 

7

MCNPX- Where can you get it? 

� Distributed by Radiation Safety Information Computational Centre 

(RSICC), Oak Ridge, US 

� http://rsicc.ornl.gov/CustomerService.aspx 

� Executables, source code and data are available 

� Cost- Free?(https://laws.lanl.gov/vhosts/mcnp.lanl.gov/ )- check 

with RSICC 

� Distribution is restricted –subjected to export control 

8

MCNPX- Advantages 

� Well established Monte Carlo code 

� Accurate model of particle transport 

� Requires no programming knowledge ..!☺ 

� Initial learning curve is not steep 

� Can be used for large range of energies 

� Single input data file 

9

MCNPX- Disadvantages 

� Interface may not be great, especially for complex problems 

� Geometry input is difficult by the requirement to refer to objects 

by number 

� Basic text output- input deck is very specific 

� Distribution of the code is restricted 

10

MCNPX input file deck 

11

MCNPX- Input file 

� User generated input file is read by MCNPX 

� Single input file 

� Format of input file is very specific 

� File consists of information on: 

� Radiation sources 

� Material description 

� Geometry of the problem 

� Types of results (tallies) required from simulation etc. 

12


� Basic units used in MCNPX 

� Length-cm 

� Energy- MeV 

� Time- sec 

� Temperature- kT 

� Atomic density- atoms/barn-cm 

� Mass density- g/cm 3 

� Cross section- barns (10- 24 cm 2 ) 

13


� Input file consists of 3 major sections 

1. Cell cards- define the shape and material content of physical space of the problem 

2. Surface Cards –defines the boundaries in space used to create cells (cylinder, sphere etc.) 

3. Data cards – defines problem type, source specification, materials, tally specification etc. 

� “Card” describes single line of input that can consists of up to 80 character 

� A “sections” consists of one or more cards 

14


General Format 

Message block 

Blank line Delimiter 

Title card 

Cell card block 

. 

. 

Blank line delimiter 

Surface Card block 

. 

. 

Bank Line Delimiter 

Data Card block 

. 

. 

optional 

� MCNPX interprets blank line as the end of preceding information block 

� $ sign terminate data entry, anything follows is interpreted as a comment 

� If the first five columns of a card are blank, the entries on the card are 

interpreted as a continuation of the data from he last named card 

15

MCNPX- Input file – Cell cards 

� Cells are basic geometry unit 

� Cartesian co-ordinate system 

� Cells are defined as volumes of space bounded by surfaces 

� Surfaces are used to define space by combining 

� Sign defines surface ‘sense’ (+/-) 

� Variable operators- Intersections (‘space’), Union (‘:’) or complement (‘#’) 

16

MCNPX- Input file- Cell Cards 

� Cell cards- define the shape and material content of physical space of the problem 

� The specific cell card format 

j m d geom params 

j -cell number, an integer from 1 to 999999 

m - material present in a particular cell, an integer from 1 to 99999 or 0 if the cell is a 

void 

d - cell material density, a positive entry is interpreted as atomic density in units of 

atoms/barn-cm, a negative entry is interpreted as the mass density in units of g/cm 3 . 

geom – specification of geometry of the cell. Boolean operators in conjunction with 

signed surface numbers to describe how the regions bounded by the surfaces 

are to be combined. 

params- optional importance card 

17


� Cell card format example 

2 1 -1 -1 imp:p=1 

2 �� cell number 

1 �� material number 

-1 �� density (in g/cc) 

-1 �� cell 2 is bounded only by surface 1. surface 1 will be defined in the 

surface card section. Negative sign indicates that cell 2 has a negative 

sense with respect to surface 1. 

imp:p=1 �� importance card for protons. imp:p=0 will terminate particle, normally 

used outside simulation cell 

18


� Cell card format-description of cells -another example 

1 +1 -2 +3 -4 

2 #1 

#1 means that the description of the current cell is the complement of the description 

of cell 1. 

y 

surface 1 

X 

surface 4 

Cell I 

surface 3 

surface 2 

Cell II 

19

MCNPX- Input file- Surface cards 

� Surface card- defines the boundaries in space used to create cells 

� Surface card – general format 

j a list 

j �� surface number (1-99999) 

a ��surface type (e.g. cz- infinite cylinder centred on z axis,) 

list �� number that describe the surface (e.g. radius of a sphere) 

20


� surface card format- example 

100 so 60 

100 �� surface number 

so �� sphere centred at origin 

60 �� 60cm radius 

21


22

MCNPX- Input file- Data cards 

� Data Cards- defines problem type, source specification, materials, tally specification etc. 

� Format is similar to cell card and surface card 

� Some important data cards relevant to medical physics include: 

• Material card 

• MODE card 

• Importance card 

• Source 

• Tallies 

• Cut-off’s etc. 

23

MCNPX- Input file- Data cards-Material card 

� Material Card format 

mn ZAID1 fraction1 ZAID2 fraction2 

mn �� material card number (m) followed by the material number 

ZAID �� element identifier, ZZZAAA where ZZZ- atomic number, AAA-mass number 

fraction �� fraction of the material/nuclide. Positive value indicates atomic fraction, 

negative indicates weight fraction 

24

MCNPX- Input file- Data cards-Material card 

� Material Card example- water (H 2 O) 

m1 1001 2 8016 1 

m1 �� material number 

1001 �� Hydrogen 

2�� atomic fraction 

8016 �� Oxygen 

1 �� atomic fraction 

25

MCNPX-Input file- Data cards- Source definition 

� General format 

SDEF source variable=specification 

SDEF �� most commonly used in medical physics calculations. Often 

we use other three cards with SDEF, Source Information (SI), 

Source probability (SP) and Source Bias (SB) cards. 

source variable�� commonly used variables- CEL(cell number),SUR (surface 

number), ERG(energy-MeV), POS(reference point ),PAR 

(particle type from source),X-x coordinate of position, AXS- 

(reference vector) etc..(TABLE 5.5.1 in manual) 

26

MCNPX-Input file- Data cards- Source definition 

� source definition card example- Isotropic disc source of radius 5 cm 

emitting 1.2MeV photons 

SDEF par=2 erg=1.2 sur=101 pos=0 0 0 rad=d1 

SI H 0 5 

SP1 -21 1 

par 2, erg=1.2 ��particle type photons, energy 1.2 MeV 

sur �� surface number on which the source is placed 

pos �� position of the centre of the disc 

rad=d1 �� radius of the source is defined as a distribution 

SI H 0 5 �� H indicates the source distribution is a histogram 

SP1 ��card distribute the source evenly across the disc surface using -21,which is a 

built in power law function 

27

MCNPX- Data cards- Source definition 

28

MCNPX- Input-Data cards- MODE card 

� MODE card specifies which particle should be created and tracked 

MODE p1 1 ,…p1 i 

P1��particle designator. 

e.g. MODE N H E will transport neutrons, protons and electrons 

29

MCNPX- Input-Data cards- PHYS card 

� PHYS- Particle physics options. Proton physics format 

PHYS:H emax ean tabl J istrg J recl J J J efac 

emax �� maximum proton energy limit 

ean �� analogue energy limit, 

tabl �� table based physics cut-off, 

J��unused place holder 

istrg �� controls charged particle straggling 

recl �� light ion and heavy ion recoil control account for the ionisation potential 

efac �� controls stopping power energy spacing 

e.g. PHYS:H 100 0 -1 J 0 J 0 3J 0.917 

30

MCNPX- Input- Data cards-Tally specification 

� Tally card- specify what type of information user wants to gain from the 

calculation . e.g. flux at a point, average energy deposition over a cell, etc. 

� General format 

Fn p1 s1,s2 …., T 

n�� tally number. 

p1�� particle designator- N,P, E 

s1 �� problem number of surface or cell for tallying 

T �� total over specified surface or cells for F1and F2 tallies 

31

MCNPX- Input- Data cards-Tally specification 

• Tally card- example 

F2:N 1 3 6 T 

• F2 is the flux averaged over a surface 

• Above card specifies four neutron flux tallies, one across each of the 

surfaces 1,3, 6 and one which is the average of the flux across all three of 

the surface, T 

32

MCNPX-Input- Data cards- Tally specification 

� TMESH- The Mesh Tally general format 

TMESH CORAn CORBn CORCn ERGSH MSGMF RMESH CMESH SMESH ENDMD 

• TMESH- Graphically displaying particle flux, dose or other quantity on a 

rectangular, cylindrical or spherical grid overlaid on top of the standard 

problem geometry 

• Mesh tally data are written to the MDATA file at the end of each run, which 

can be converted into a number of standard formats suitable for reading by 

various graphical analysis package 

• CORAn,CORBn and CORCn cards describe a mesh in three co-ordinate 

direction as defined by mesh type e.g. CORA3 0 10i 20 for a 

rectangular mesh, the values define planes perpendicular to the x-axis. i 

enables a large number of regularly spaced mesh points to be defined with 

minimum entries 

• ERGSHn e1 e2- lower and upper energy limits for information to be 

stored to mesh tally n. Default –all energies 

• RMESH- rectangular mesh, CMESH- cylindrical mesh, SMESH- Spherical 

mesh 

• ENDMD- terminate mesh tally 

33

MCNPX- Processing Mesh Tally Results 

� Mesh tallies may be plotted with MCNPX geometry plotter. 

• MCNPX Z 

• MCPLOT>RUNTPE=filename 

• MCPLOT 

• PLOT>py 4 ex 40 or 0 4 0 la 0 0 tall2 color on la 0 0 con 0 100% 

� Or use MCNPX tally plotter 

• MCNPX Z 

• MCPLOT>RMCTAL=filename 

• Tal 12 free ik 

� Or post process the MDATA (MCTAL) file with GRIDCONV and then use and external 

package. GRIDCONV converts the data arrays in MDATA to form compatible with 

various external graphics package e.g. IDL,GNUPLOT, PAW etc.. 

A useful programme: A Matlab code to import the mesh tally information contained in an MCNPX 

mdata file (http://www.ptcri.ox.ac.uk/people/DanielWarren/ReadMData.shtml ) 

34

MCNPX installation 

35

MCNPX- installation and running the code 

� Installation is simple for executable files 

� For Windows, unzip the files, create folders, place the files in the right folders… 

(Manual Sec 3- for details and other platforms) 

� Download and install an X-Server or similar software for graphics 

� MCNPX Execution line; 

� MCNPX KEYWORD=value….KEYWORD=value execution_option 

other_options 

� KEYWORD- MCNPX default filename. User may assign a specific value (file name or 

path), execution_option is a character or string of character informs MCNPX which 

of five execution modules to run, other_ option provide the user with additional 

execution control 

More Info: User Manual or http://mcnpx.lanl.gov/opendocs/installation/README.w32 

36

MCNPX- Installation and running the code 

37

MCNPX- Related software 

� The Visual Editor for MCNPX 

� (http://www.mcnpvised.com/mcnpx/mcnpx.html) 

� Included with MCNPX 2.6 or higher. 

� Can be used to display, surface creation/cell creation etc. 

� Some detailed examples can be found in 

http://www.mcnpvised.com/sample_exercises/exercises/create.pdf 

38

MCNPX- Related software 

� MORITZ(http://www.whiterockscience.com/moritz.html) 

� Interactive geometry editor/view for MCNPX 

� Read output from MCNPX and can visualise geometry and particle tracks in 

2D and 3D,write input files 

� SABRINA (http://www.whiterockscience.com/sabrina/sabrina.html) 

� Scan2MCNP (http://www.whiterockscience.com/scan2mcnp/Scan2MCNP.html ) 

converts CR,MRI and other scan data to an input file 

� Codes may not be free. 

39

MCNP/MCNPX- Some useful literature 

1. MCNPX user manual- http://mcnpx.lanl.gov/documents.html 

2. Medical Physics Calculations with MCNP: A primer, Masters Thesis by Alexis Reed 

(http://mcnp-green.lanl.gov/pdf_files/la-ur-07-4133_mp-primer.pdf )- dose from 99m Tc 

therapy, calculation of Medical Internal Radiation Dose (MIRD) specific absorbed fraction values using 

ORNL MIRD phantom, x-ray phototherapy effectiveness, prostrate life time dose calculations, and 

radiograph of the head using the ZUBAL head phantom 

3. An MCNP primer, J K Shultis et 

al.,(http://web.utk.edu/~rpevey/public/NE406/MCNP%20primer.pdf ) 

4. Varian 2100C photon beam and dose delivery ,Masters thesis by Jared 

Hoover(http://smartech.gatech.edu/jspui/bitstream/1853/16245/1/hoover_jared_s_20070 

8_mast.pdf ) 

5. Voxel phantom setup in MCNPX by V Taranenko, M Zankl, H Schlattl 

6. MCNPX network (http://mcnpx.net/ ) 

40

MCNPX –Applications in particle therapy 

41

MCNPX- Applications in particle therapy 

� Large increase in particle therapy centres world wide 

� Particle therapy is still not completely optimised 

� Uncertainties in particle range, especially for tissue inhomogenity and interface 

� MCNPX has got wide applications in particle therapy 

� Can be used for dose deposition, accurate range estimation 

� Useful for Carbon ion therapy calculation also 

42

MCNPX- Proton pencil beam simulation 

A sample input deck- Proton pencil beam dose deposition in water 

1- 

2- 

3- Proton pencil beam simulation in water $optional title card 

4- C 

5- C *********************************************************** 

6- C CELL CARDS 

7- C *********************************************************** 

8- 1 1 -1 -1 IMP:P=1 $cell num, material num, density, surface no, relative cell importance 

9- 3 0 -99 1 IMP:P=1 $ $bounding box for problem - cell 3 - void 

10- 

11- 

99 0 99 IMP:P=0 $ void outside bounding box 

12- C *********************************************************** 

13- C SURFACE CARDS 

14- C *********************************************************** 

15- 1 RPP 0 30 -8 8 -8 8 $RPP-rectangular parallel piped, RPP Xmin Xmax Ymin Ymax Zmin Zmax 

16- 

17- 

99 RPP -15 45 -16 16 -16 16 

18- C ********************************************************** 

19- C DATA CARDS 

20- C ********************************************************** 

21- C ------ MATERIAL CARDS ------ 

22- M1 1001 2 8016 1 $ 1001-hydrogen, 2 atomic fraction, 8016-oxygen 

23- C 

24- C 

25- C --------MODE CARD----- 

26- MODE H E P $ specifies what particles might be created and tracked 

27- CUT:H J 1 $ cut off value for proton 

28- CUT:E J 1 $ cut off value for electron 

29- CUT:P J 1 $ cut off value for photon 

30- PHYS:H 400 0 -1 J 0 J 0 $ proton physics format 

43


A sample input deck 

31- C 

32- C ------- SOURCE DEFINITION------ 

33- SDEF X=-5 AXS=1 0 0 EXT=0 Y=d1 Z=d2 DIR=1 VEC=1 0 0 PAR=H ERG=200 $SDEF source definition 

34- SP1 -41 1.1774 0 $distribution uses FWHM = 2root(2*ln2) of sigma.SP1-source probability,-41 

is for Gaussian. distribution uses 0.70645 width of FWHM, 0 is the mean, 

FWHM = 2root(2*ln2) of sigma, X-x coordinate of position, AXS-reference 

vector, refer Table 5.5.1 in manual 

35- SP2 -41 1.1774 0 $distribution uses FWHM = 2root(2*ln2) of sigma 

36- C 

37- C ------MESH TALLY CARDS------- 

38- TMESH $TMESH- Energy Deposition Mesh Tally 

39- RMESH3 $RMESH is a Rectangular mesh 

40- CORA3 0 80i 20 $planes perpendicular to the x-axis 

41- CORB3 -8 64i 8 

42- CORC3 -8 64i 8 

43- ENDMD $end mesh tally 

44- C 

45- C --------NUMBER OF PARTICLE HISTORIES TO RUN---- 

46- NPS 1000000 $num of particle histories 

47- PRINT 10 110 170 $output print tables format PRINT x1,x2,..i.., xi-lists of table numbers 

to be included in the output file. 10-source co-efficient and 

distribution, 110-first 50 starting histories, 170-source distribution 

frequency table, surface source 

44


Output plot using Matlab 

45

MCNPX- Treatment planning calculations for hadron 

therapy 

� MCNPX can be used to develop a treatment planning software 

� Steps involved: 

� From the CT image, estimate HU values for each voxel 

� Convert HU values to corresponding chemical composition using a 

standard calibration graph 

Schneider, U., E. Pedroni, and A. Lomax, 1996, “The calibration of CT Hounsfield units for radiotherapy 

treatment planning,” Phys. Med. Biol. 41, 111–124. 

46

MCNPX- Treatment planning calculations for hadron 

therapy 

� Steps involved: 

� Need some software to automatically create the MCNPX input file 

(Matlab, IDL etc..) 

� Run the simulation 

� Automatically read plot results and superimposed on CT image 

Useful software: CERR(http://www.cerr.info/about.php )- Freeware: written in Matlab to 

import and display treatment plans from wide variety of commercial or academic TPS 

including RTCOG and DICOM-RT format 

47

Happy particle crunching.......!!!!! 

Milton Woo 

Eva Bezak 

John Thomas 

Daniel Warren 

Thank You 

48

MCNPX Wilson

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