Use of tally with surfaces and macrobodies in MCNP

In summary: So if you wanted to include the entire 600 px area, you would need to do an f1 or f2 over the entire box.
  • #1
Geovanny Gutierrez
5
0
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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  • #2
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.
 

Related to Use of tally with surfaces and macrobodies in MCNP

1. What is MCNP and how is it used?

MCNP stands for Monte Carlo N-Particle Transport Code, and it is a computer program used to simulate the transport of particles through matter. It is commonly used in nuclear engineering and medical physics to model radiation and neutron transport. MCNP uses a combination of physics models and statistical methods to calculate the behavior of particles as they interact with matter.

2. How does tallying work in MCNP?

Tallying is the process of recording and analyzing the results of particle transport simulations in MCNP. Tallying involves specifying a set of particles, energies, and locations to be tracked and recording the number of particles that reach those locations. This information can then be used to calculate quantities such as dose, flux, and reaction rates.

3. What are surfaces and macrobodies in MCNP?

In MCNP, surfaces are geometric boundaries that define the shape and size of objects in a simulation. They can be simple shapes like planes, spheres, or cylinders, or more complex shapes defined by mathematical equations. Macrobodies are collections of surfaces that are used to represent larger objects, like buildings or vehicles. They can be composed of multiple surfaces and can have a variety of properties, such as material composition and density.

4. How are surfaces and macrobodies used in MCNP?

Surfaces and macrobodies are used in MCNP to define the geometry of a simulation. By specifying the shape, size, and material properties of surfaces and macrobodies, scientists can accurately model the behavior of particles as they interact with different materials. This allows for realistic simulations of complex systems, such as nuclear reactors or medical imaging devices.

5. What are the benefits of using tally with surfaces and macrobodies in MCNP?

The use of tally with surfaces and macrobodies in MCNP allows for accurate and efficient simulations of particle transport through complex systems. By defining the geometry of a simulation with surfaces and macrobodies, scientists can obtain precise results for quantities of interest, such as dose and flux. This can aid in the design and optimization of various systems, such as radiation shielding or medical imaging devices.

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