MCNP Surface Tallies: F1 & F2 on Infinite Cylinders & Planes

In summary: No, that's not what the union means. The union means that the surface of the cell is the set of points that are in the union of the surfaces.
  • #1
19matthew89
47
12
TL;DR Summary
How to define surface tallies of a cell?
Hi,
I have a question concerning surface tallies like F1 and F2. You have to provide a surface for them. Since, surfaces are defined as infinite (infinitely long cylinders, infinitely extended planes) how can you write the surface tally of a cell? What are the actual tally surfaces for F1 anf F2? The whole (infinite) surfaces or the actual surface of a cell?

As a quick example let's consider a z cylinder

So

Code:
1 M# rho_M -2 3 -4

2 CZ 1
3 PZ -10
4 PZ 10

will define the cell 1 which is a cylinder of radius 1 cm and height 20 cm.

May aim is tallying the surface flux F2 through the surface of the cylinder. How can I do it?

Is this
Code:
F2:N 2 3 4 T

right?

The manual says that if I use the parenthesis I will get a tally through the union of the surfaces but I'm not interested in the union.

Thanks a lot in advance!

P.S. I know a solution to be sure would be segmenting, but I am hoping for a less cumbersome solution and also then about what F2 (or F1) are actually evaluating.
 
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  • #2
Hi,
you can with a macrobody surface "RCC"
In your case :
c cell
1 M# rho_M -10

c surface
10 RCC 0 0 -10 0 0 20 1

f2:n 10
 
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  • #3
PSRB191921 said:
Hi,
you can with a macrobody surface "RCC"
In your case :
c cell
1 M# rho_M -10

c surface
10 RCC 0 0 -10 0 0 20 1

f2:n 10
Hi, thanks!

Indeed that could work, but but:
1. my supervisor strongly recommended that I don't use macrobodies. He says that, according to his experience, the code behaves "weirdly" with macrobodies. I don't know what he means with that but I'll try to stick to hi suggestion and not using macrobody structures.
2. I still want to be able to have the tallies on the individual three surfaces defining the cylinder (bottom end, top end and lateral cylindrical surface)
3. Still it's not that clear to me what the tally F2:N 2 3 4 T then does. Clearly you have to assign a surface area via SD card otherwise it cannot compute the tally. Do you confirm then that F2/F1 consider the whole infinite surface for the tally computation? Thanks!
 
  • #4
Wow, what an unexpected nightmare of a question. Coool.

19matthew89 said:
TL;DR Summary: How to define surface tallies of a cell?

The manual says that if I use the parenthesis I will get a tally through the union of the surfaces
This is exactly what I expected, trouble is now I check the manuals this isn't what they say. They say this produces an average flux. Do us the favour of double checking, cheers.

The surfaces are not infinite practically because nothing is tracked through the void.

"F2:N 2 3 4 T"
This should be 4 tallies, Surface 2, 3, 4, and the average of the three surfaces.
 
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  • #5
Alex A said:
Wow, what an unexpected nightmare of a question. Coool.

This is exactly what I expected, trouble is now I check the manuals this isn't what they say. They say this produces an average flux.
Alex A said:
Do us the favour of double checking, cheers.
I'm afraid I don't understand your point. It indeed produces the average flux of the union of the surfaces. But that is not useful in defining the surface I'm interested in.
Alex A said:
The surfaces are not infinite practically because nothing is tracked through the void.

"F2:N 2 3 4 T"
This should be 4 tallies, Surface 2, 3, 4, and the average of the three surfaces.
Yeah. It is exactly what it does but, from what I understand, it takes into account all the particle crossing the surfaces "at the infinite", namely:
* for the cylinder, the particle crossing the cylinder above the plane defined by surface 4 (@10 cm) and below the plane defined by surface 3 (@-10 cm)
* for the planes 3 and 4, the particles crossing the planes outside the circle defined by the inner part of surface 2

In other words it seems that there is no way of defining just the surface of I'm interested in but using macrobodies.

By the way I "solved" it by segmenting the infinite surfaces and using SD card.
For instance for surface of the circle defined by the intersection of plane 3 and cylinder 2, I wrote

Code:
F2:N 3

FS2 -2

SD2 3.141592 1

and then I'll only consider the tally of the first segment.

Finally a last comment about the solution of PSRB191921. I tried that solution out and, as pure simple example, it worked but remember that MCNP does not accept simply this (at least my code was complaining).
PSRB191921 said:
c surface
10 RCC 0 0 -10 0 0 20 1

f2:n 10
You have to specify the facets of the macrobodies for the tally F2 to work. So namely you have to write

Code:
10 RCC 0 0 -10 0 0 20 1f2:n 10.1 10.2 10.3 T

if you want to have the tallies on the three surfaces defining the cylinder and an average on it.

Cheers!
 
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  • #6
I assumed union meant as in constructive solid geometry. The manual does use the word "union" and does clarify it means "average" for the F2 tally. You are right!, and you are also right you cannot specify an area this way.

FS is the right way to solve it, but it's not obvious what it means when a surface is subdivided by more than one other surface! The result is tree like.
 
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FAQ: MCNP Surface Tallies: F1 & F2 on Infinite Cylinders & Planes

What is the difference between F1 and F2 tallies in MCNP?

The F1 tally in MCNP measures the flux of particles crossing a surface, essentially counting the number of particles per unit area that intersect the surface. The F2 tally, on the other hand, measures the flux of particles per unit length along a surface, typically used for cylindrical or linear geometries. This makes F1 suitable for planar surfaces and F2 for cylindrical or linear surfaces.

How do you define an infinite cylinder in MCNP for surface tallies?

In MCNP, an infinite cylinder can be defined using the 'c/z' or 'c/y' surface types, depending on the orientation. For example, 'c/z' defines a cylinder with its axis along the z-axis and 'c/y' along the y-axis. You need to specify the radius and the coordinates of the center of the base of the cylinder. For instance, 'c/z 0 0 10' defines an infinite cylinder along the z-axis with a radius of 10 units centered at the origin.

How can you apply an F1 tally to an infinite plane in MCNP?

To apply an F1 tally to an infinite plane in MCNP, you define the plane using the appropriate surface card (e.g., 'pz' for a plane perpendicular to the z-axis). You then specify the tally using the F1 card followed by the surface number. For example, if you have defined a plane at z=0 using 'pz 0', you would use 'F1:n 1' to tally neutrons crossing this plane, where '1' is the surface number assigned to 'pz 0'.

What are common mistakes to avoid when setting up F2 tallies on infinite cylinders in MCNP?

Common mistakes include incorrectly defining the cylinder's orientation or radius, not properly assigning the surface number in the tally card, and misunderstanding the units of the tally results. Ensure that the cylinder's axis and radius are correctly specified and that the tally card references the correct surface number. Additionally, remember that F2 tallies give results in terms of flux per unit length, not per unit area, which is a common source of confusion.

How do you interpret the results of F1 and F2 tallies on infinite geometries in MCNP?

Interpreting F1 and F2 tally results involves understanding the units and physical meaning of the flux measurements. F1 tally results are given in terms of particles per unit area crossing a surface, useful for understanding the particle distribution across a plane. F2 tally results are given in terms of particles per unit length along a cylinder or line, which is helpful for analyzing flux along cylindrical geometries. Ensure you convert these results appropriately

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