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

[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

 

[DEvens]

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.

 

[Deb]

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)?
Thanks again for your help