# MCNP6.2: F2 tally and problem with the area calculation

• Deb
In summary, MCNP calculates the particle flux through the two surfaces used to close the concrete box.
Deb
TL;DR Summary
I woud like to calcultate a dose (with an F2 tally) through a surface with MCNP6.2 and the code cannot calcute the area of the surface.
Hello,
after many simplifications my geometry has become very simple: just a box of concrete with a cylinder of steel inside. The source is outside in the air. The cell and the surface cards are like the following:

C ******************Block 1: Cells**********************
100 0 99 imp:p=0
99 1 -0.00122 (-99 2.3):( -99 2.2) imp:p=1 $Air 1 2 -2.3 -99 2 -2.3 -2.2 imp:p=1$ Concrete
2 3 -7.6 -2 imp:p=1 \$ Cylinder (steel)

C *******************Block 2: Surfaces**************************
99 RPP 0.0 450.0 50.0 172.0 50.0 172.0
2 RCC 270.0 111.0 111.0 140.0 0.0 0.0 4.5

I attach a picture of the geometry.

In the output file I have this message:
surface input calculated reason area
area area not calculated
2 99.1 0.00000E+00 1.48840E+04
3 99.2 0.00000E+00 1.48840E+04
4 99.3 0.00000E+00 5.49000E+04
5 99.4 0.00000E+00 5.49000E+04
6 99.5 0.00000E+00 5.49000E+04
7 99.6 0.00000E+00 5.49000E+04
9 2.1 0.00000E+00 3.95841E+03
10 2.2 0.00000E+00 0.00000E+00 asymmetric
11 2.3 0.00000E+00 0.00000E+00 asymmetric

I guess that the areas of the surfaces 2.2 2.3 are not calculated because they are used both as basis for the cylinder and for closing the concrete box. Therefore MCNP cannot distinguish which size he has to assign to the surfaces.
I want to calculate the dose (with an F2 tally) at the entrance and at the exit of the concrete, which means through the surface 22.2 and 22.3, whose area is not automatically calculated. I could use the SD card to pass the correct area to the code.
-My first question is: the particle flux is calculated through the two basis of the cylinder or through the two rectangles closing the concrete box?
-Second question: is there a way to solve this problem? I have tried also using two the px planes to close the concrete cell, but this does not help, since MCNP cancel the identical surfaces.

Thanks to anyone, who can help me.

Debora

#### Attachments

• Geo.PNG
1.3 KB · Views: 368
Can you post a full file that MCNP will run?

When a tally needs an area, and MCNP can't calculate it, you can provide it using the SD card. Quoting from the manual: MCNP6 cannot calculate the volumes and areas of asymmetric, non-polyhedral, or infinite cells.

If you and MCNP are both unable to calculate the area, then you can use the stochastic volume and area calculation method. I have currently got MCNP 6.1, so maybe your user manual is different to mine. But for me it's in Section 3.3.1.1 Cell Volume.

Thanks for your quick answer. I attach the full input file now.
It´s clear to me that I can solve with the SD card,
but since the geometry is so easy, my doubt is why MCNP doesn´t calculate the area. I can calulate it, sure.
My major doubt is through which surface does it calculate the flux? The area is used only for the normalization, but if I say
f2:p 2.3
is it really the flux trought the entrance concrete+cylinder (to be precise area=14884 cm^2)?
and not only trought the basis of the cylinder (area 54,08cm^2)?

#### Attachments

• SoloCemento.txt
2.1 KB · Views: 366

## 1. What is MCNP6.2?

MCNP6.2 is a Monte Carlo N-Particle transport code used for simulating and analyzing nuclear systems and radiation transport. It is widely used in the nuclear industry and research fields.

## 2. What is the F2 tally in MCNP6.2?

The F2 tally in MCNP6.2 is a type of scoring method used to calculate the flux of particles passing through a specified surface or volume in a simulation. It is commonly used to analyze neutron and photon transport in nuclear systems.

## 3. What is the problem with the area calculation in MCNP6.2?

The problem with the area calculation in MCNP6.2 is that it can lead to inaccurate results if the geometry of the simulation is not properly defined. This can occur when the surface or volume used for the F2 tally is not well-defined or when there are overlapping surfaces or volumes in the simulation.

## 4. How can the problem with area calculation be resolved in MCNP6.2?

The problem with area calculation in MCNP6.2 can be resolved by carefully defining the geometry of the simulation and avoiding overlapping surfaces or volumes. It is also important to verify the results by comparing them with other scoring methods or experimental data.

## 5. Are there any alternative methods to the F2 tally in MCNP6.2?

Yes, there are alternative methods to the F2 tally in MCNP6.2 such as the F4 tally, which calculates the flux on a specified surface or volume and also takes into account the direction of the particles. Other scoring methods include the track-length tally and the point detector tally.

• Nuclear Engineering
Replies
2
Views
2K
• Nuclear Engineering
Replies
58
Views
3K
• Nuclear Engineering
Replies
1
Views
188
• Nuclear Engineering
Replies
1
Views
2K
• Nuclear Engineering
Replies
5
Views
2K
• Nuclear Engineering
Replies
3
Views
2K
• Nuclear Engineering
Replies
4
Views
2K
• Nuclear Engineering
Replies
0
Views
2K
• Nuclear Engineering
Replies
1
Views
2K
• Nuclear Engineering
Replies
1
Views
3K