MCNP4 help: f4 tally in lattice

[J_P_C]

TL;DR

I can't find the way to correctly apply an f4 tally in a precise lattice position. Help please?

Hi all, I'm new to the forum. Maybe you guys can give me hand with this.

I am using MCNP4 to model a 17x17 fuel element. I want to know the average neutron flux in a specific pincell but so far everything I try results in error. This is my input (text file is attached too):

c CELL CARDS
1  1 -10.24  -7        u=1     imp:n=1         $Fuel
2  2 -6.504   7 -8    u=1     imp:n=1         $Clad
3  3 -1.000   8         u=1     imp:n=1         $Water outside clad
4  3 -1.00   -9         u=2     imp:n=1         $Water inside guide tube
5  2 -6.504   9 -10  u=2     imp:n=1         $Guide tube
6  3 -1.000   10       u=2     imp:n=1         $Water outside guide tube
7  0          3 -4 5 -6 lat=1  u=3  imp:n=1  $Fuel element lattice
              fill=-8:8 -8:8 0:0
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1
              1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1
              1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
              1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
8  0             11 -12 13 -14           fill=3  u=4  imp:n=1  $Fuel element
9  3   -1.0  -11:12:-13:14                        u=4  imp:n=1  $Water outside fuel element
10  0            1 -2 15 -16 17 -18  fill=4            imp:n=1  $Fuel element + water gap
999 0          -1:2:-15:16:-17:18                        imp:n=0  $Outside void

c SURFACE CARDS
*1  pz -200
*2  pz  200
 3  px -0.6325
 4  px  0.6325
 5  py -0.6325
 6  py  0.6325
 7  cz  0.4127 
 8  cz  0.4744
 9  cz 0.57294
 10 cz 0.61301
 11 px -10.7525
 12 px 10.7525
 13 py -10.7525 
 14 py 10.7525
*15 px -10.8055
*16 px 10.8055
*17 py -10.8055
*18 py 10.8055

I want to know the average neutron flux in the top right pincell.
I don't understand why the following tally results in the fatal error "tally volume or area was not input nor calculated":

f4:n ((1 2 3) < 7 [8 8 0] < 8 <10)

The way I see it all of my cells are finite when filled with their corresponding universes. I'm a bit lost right now. Any help is appreciated.

 

[Alex A]

F4 is a flux tally, but instead of counting the neutrons going through it MCNP measures the total path length and divides it by the volume of the cell to calculate the flux. It's only the cookie cutting later that gives (1 2 3) finite volume. So the function to find the volume fails and MCNP freaks.

Fixed by specifying the divisor explicitly, which in this case is the volume. Adding "SD4 1" the line after the F4 tally causes it to run, either replace 1 with the real volume, do the division later or rewrite the tally cells to be bounded (the problem is neatly defined so I imagine you don't want to go this way).

 

[J_P_C]

F4 is a flux tally, but instead of counting the neutrons going through it MCNP measures the total path length and divides it by the volume of the cell to calculate the flux. It's only the cookie cutting later that gives (1 2 3) finite volume. So the function to find the volume fails and MCNP freaks.

Fixed by specifying the divisor explicitly, which in this case is the volume. Adding "SD4 1" the line after the F4 tally causes it to run, either replace 1 with the real volume, do the division later or rewrite the tally cells to be bounded (the problem is neatly defined so I imagine you don't want to go this way).

Thanks for the answer, I see what you mean. In fact, now that I know I need the sdn card I went looking into the manual and I saw that this card is especially recommended for tallies in repeated structures. I liked the idea of using the divisor just as 1 and dividing by the real value later, so I added the "sd4 1" line after the tally line, but still I get the same error. Maybe the chain of cells in the f4 tally is not correct? I am attaching the full, complete input to this message too.

 

[Alex A]

Your input file works unmodified for me with 4b under linux, I just renamed it to inp. I've run it with 4c but the binary was older and I needed to strip carriage returns (a linux thing). I assume you are in windows.

Try reading the output file to make sure the modification is in the input it is processing, for whatever reason. Check for any other errors or warnings. If it still doesn't work post the output file.

 

[J_P_C]

It finally works, although I have no idea why . I ran the very same input file from my previous message and this time it worked. And yes, I am in Windows. Thanks so much for the help!

My original intention was to compare the flux in an innmermost pincell versus flux in an outermost pincell. If by any chance anyone has my same problem in the future and finds the thread, the file I am attaching to this message is working and does what I intended.

Thanks again for the help.