MCNP: Declaring two sources in two cells

  • Thread starter lee phong
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In summary, "MCNP: Declaring two sources in two cells" discusses how to define multiple radiation sources within different cells in the Monte Carlo N-Particle Transport Code (MCNP). The document provides guidance on specifying source parameters for each cell, ensuring accurate simulations of particle behavior and interactions in complex geometries. It emphasizes the importance of correct syntax and source definitions to achieve reliable computational results in radiation transport modeling.
  • #1
lee phong
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Dear all,
I am a new user of MCNP, I have a problem declaring 2 sources F-18 and I-131 in 2 different cells. I only know how to declare each source for each cell, but when combining them into one declaration, I don't know how to do it. Here are my two source declarations. Please help me. Thank
 

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  • #2
Welcome to physicsforums lee phong!

I don't think there is a real disadvantage to doing two runs with different sources and combining the results. This also has the advantage that some mistakes are easier to spot and if you want to change the source strength of one relative to the other you don't need to rerun the simulation.

If you wrote, or fully understand the F-18 sdef, then you understand everything you need to extend this to a second source. The file uses dependent variables to emit photons, with one probability distribution (d12) and electrons with another (d13), the iodine is a photon source you can rename this to make d14 and add those lines. Then extending SI11 to be L P E P, extend SP11 for the strength. And extend DS10 to DS10 S 12 13 14

Does that make sense? If the iodine was an electron source too it might be SI11 L P E P E, and then all dependent variables of PAR would need 4 entries.

With three entries, you need a dependent variable for POS. DS15 L ax ay az bx by bz cx cy cz
Where a is the position for the F-18 photons, b is the position of the F-18 electrons (the same) and c is the location for the I-131 photon source. In sdef pos becomes pos=fpar=d15 from the default 0 0 0.

You may need another dependent variable for CEL, since the sources seem to be in different cells but it will work exactly the same way.

If anything isn't clear, just ask.
 
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  • #3
Thank you, for your guidance. This my declaring for the sources:
SDEF PAR= D1 ERG = FPAR = D2 CEL = D3
SI1 L P E P
SP1 1 9.67300E-01 1.93460E+00 $Probality of I-131 & F-18, respective
DS2 S 10 11 12
SI10 L 0.029458 0.029779 0.284305 0.364489 0.636989 0.722911$Photon energy of I-131
SP10 D 0.0138 0.0256 0.0614 0.817 0.0717 0.0177 $ Probality of I-131
SI12 L 5.11000E-01 $Photon energy of F-18
SP12 D 1 $ Prob of photon F-18
SI11 L 2.495E-01 $Mean of electron energy F-18
SP11 D 1 $Prob of electron energy F-18
SI3 L 2 3 $Cell discrible
SP3 D 1 1 $Prob of each cell
This is my describe of the sources. Is it right ? But my outp file noticed "fatal error" with "d1" and "p" in SI1 L P E P.
Please help me
 

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  • #4
CEL=D3 needs to be a dependent variable, you also don't have a POS yet. SI1 L P E P, is the same as SI1 L 2 3 2. But don't bother to fix this, it can't work in mcnp5. Setting PAR to a distribution works in 6 and X but not in 5. You can have sources in different cells but only one particle type at a time.

Doing two or more runs, multiplying the tallies by the activity of that source and summing the source results is still a good plan. You will need to split the F-18 sdef.
 
  • #5
Ok, thank you. I found some problems with MCNP5. My problem with F-18 and I-131 liquid mixing in 2 tanks and each tank has a different ratio volume F-18: I-131. So I separated it into 2 runs with photon energy (insert 511keV of F-18 in I-131) source and electron source of F-18. Also, I describe POS for CEL=D3 and the probability each cell follows the ratio volume. Is that right?
 
  • #6
That is right. If you are happy to share your input file, I or people in the forum might be able to offer more advice.

If you want a better simulation you could use volume sources instead of point sources. How complicated this is may depend on the shape of your tanks, and you might need to do more runs to get the final answer.
 
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  • #7
Great thank for your support
 
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FAQ: MCNP: Declaring two sources in two cells

What is MCNP and how does it relate to declaring sources in cells?

MCNP (Monte Carlo N-Particle Transport Code) is a software package used for simulating the transport of neutrons, photons, and electrons. Declaring sources in cells allows users to specify where particles originate within a defined geometry, which is essential for accurate simulations of radiation transport in various materials and configurations.

How do I declare multiple sources in different cells in MCNP?

To declare multiple sources in different cells, you need to use the SDEF card for each source. Each SDEF card specifies the source parameters, such as type, energy, and position. You can define the sources separately for each cell by using the appropriate cell numbers in the SDEF statements, ensuring that the sources are clearly identified and properly configured.

Can I use different source types in different cells in MCNP?

Yes, you can use different source types in different cells in MCNP. Each source can be defined independently with its own parameters, allowing for a variety of source types, such as point sources, isotropic sources, or directional sources, to coexist within the same simulation. Just ensure that each source is correctly specified in its corresponding SDEF card.

What are the common errors when declaring multiple sources in MCNP?

Common errors when declaring multiple sources include incorrect cell numbers, overlapping source definitions, and improper formatting of the SDEF cards. It's crucial to double-check that each source is assigned to its intended cell and that all parameters are correctly specified to avoid simulation errors or unexpected results.

How can I verify that my sources are set up correctly in MCNP?

You can verify your sources by running a test simulation and examining the output. Check the tally results to see if the expected particle distributions and interactions occur as intended. Additionally, reviewing the input file for proper syntax and using MCNP’s built-in diagnostics can help identify any issues with your source declarations.

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