Use of tally with surfaces and macrobodies in MCNP

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SUMMARY

The discussion focuses on the use of tally cards in MCNP, specifically the f1, f2, and f4 tallies for calculating surface current and average flux on cells and surfaces. It confirms that macro bodies can be utilized with these tallies, with particular attention to the f4 tally's requirement for cell volume verification. The f1 and f2 tallies require specific surface definitions, such as using the macro body's sub-surface numbers, and emphasize the importance of understanding the positive and negative senses of surfaces within macro bodies. Users must also be aware of area calculations for these surfaces.

PREREQUISITES
  • Familiarity with MCNP (Monte Carlo N-Particle Transport Code)
  • Understanding of tally types: f1, f2, and f4
  • Knowledge of macro bodies and their surface definitions
  • Ability to calculate surface areas and verify cell volumes
NEXT STEPS
  • Review the MCNP user manual for detailed information on macro body facets
  • Learn about the calculation of cell volumes in MCNP
  • Explore the implications of positive and negative surface senses in MCNP
  • Study examples of using f1 and f2 tallies with macro bodies in practical scenarios
USEFUL FOR

Researchers, nuclear engineers, and simulation specialists working with MCNP who need to accurately model and analyze particle transport using tallies and macro bodies.

Geovanny Gutierrez
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Hi. I need some help with the use of tally card in MCNP. I have been trying to use the f1, f4 and f2 tally to calculate surface current, average flux on a cell and avergage flux on a surface respectively, my question is: It's possible use those kind of tallies with macrobodies and surfaces defined by equations at the same time?, in other words; I can use it for macrobodies and for surfaces?

Thank you.
 
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It's a bit necromantic, but this is sitting in the list of unanswered threads. After more than one year, probably it is not useful to the OP. But here goes.

Yes, you can use macro bodies with tallies.

The f4 tally is probably the most obvious. The only issue is whether MCNP is able to calculate the volume of the cell on its own. In some cases, it will get a value. You should definitely verify this value, since the details of how it does it are not directly visible to you. If it can't then you have to provide the value, since it is required to get the final flux value in particles/cm^2. The raw value reported is just the total track length counted up in the cell, per particle started.

The f1 and f2 tallies have a small wrinkle. Macro bodies allow you to specify the particular surface. Each kind of macro body assigns a number to the sub surface. For example, a box surface.

500 box -1 -1 -1 2 0 0 0 2 0 0 0 2

This has six sub-surfaces, specified by 500.1, 500.2, up to 500.6.

So if you wanted to use one of the faces of a box in an f1 tally, you provided it through this formula.

f1 500.3.

This is the third facet, which is the plane normal to the end of a2x, a2y,a2z, the "far side" of the box in the second defined dimension. In this example that turns out to be the plane y=1. Read further on in the user manual about macrobody facets.

Also, you need to be aware that the negative sense of a surface on a macro body is always the inside of the macro body. So in the box example, the surface 500.2 is a plane through x=-1. But the positive sense is outside the box, meaning negative x. That is just backward to what an ordinary px surface would be.

Also, you may need to calculate the areas for these surfaces.

Also, for macro bodies, f1 and f2 tallies are limited to the part of the surface that actually forms part of the macro body. If you defined the following surface

600 px 1

and did an f1 or f2 over it, then by default you get all of the surface x=1, over your entire model. But with a macro body as the box defined previously, 500.1 is just the portion of the surface x=1 that is part of the box.
 

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