MCNP Why does it say F5 tally is not in any cell?

In summary, MCNP may show the "F5 tally is not in any cell" error if cells have not been specified for the tally to score in. To fix this, you need to use the "F5" card before the tally card and specify the cell numbers or names. The F5 tally cannot be used without specifying cells as it needs to be placed on a surface to record data. The F5 tally is important in MCNP as it allows for recording data on a surface and can score data in multiple directions. There are alternatives to the F5 tally, such as the F4 and F8 tallies, which have different scoring capabilities and may be better suited for certain applications. It is important to understand the differences between
  • #1
llatosz
62
9
Sorry for two questions in a row, this one we have been stumped on for the entire day
We're getting "fatal error. detector no. 1 of tally 5 is not in any cell.", What could be causing this?
In the manual and all examples we've seen, nobody has parameters specifying cell location for the tally.
It is a very simple problem, 3 concentric layered cylinders on the Z axis, with a point Cf-252 source in the center, and a ring detector 1cm outside the outer surface.
Below is our simple input file if needed

1- c MCNP TEST
2- c CELL CARDS
3- 1 1 -0.96 -100 -400 500 IMP:N=1.0 $ For radius 0-10 (under plane 400, above plane 500)
4- 2 2 -11.34 100 -200 -400 500 IMP:N=1.0 $ For radius 10-20
5- 3 1 -0.96 200 -300 -400 500 IMP:N=1.0 $ For radius 20-95
6- 4 0 600 IMP:N=0 $ Terminate outside of kill-sphere
7-
8- c SURFACE CARDS
9- 100 CZ 10
10- 200 CZ 20
11- 300 CZ 95
12- 400 PZ 500 $
13- 500 PZ -500 $Two planes to cut infinitely tall cylinders
14- 600 SO 700 $Kill sphere around entire problem
15-
16- c DATA CARDS
17- MODE N
18- TOTNU
19- NPS 10000
20- SDEF POS=0 0 0 CEL=1 PAR=1 ERG=D1
21- SP1 -3 1.18000 1.03419
22- c MATERIAL SPECIFICATION
23- M1 NLIB=60c $ pOLYETHEYLENE CH2
24- 1001 2.0 $ Hydrogen
25- 6000 1.0 $ Carbon
26- M2 NLIB=60c $ Lead
27- 82207 1.0
28- c TALLY SPECIFICATIONS
29- F5Z:N 0 96 0
30- E5 1 18
31- DF5 IU=1 IC=10 FAC=6.44e7
And a snip of the output file where the error is looks like this:
28- c TALLY SPECIFICATIONS
29- F5Z:N 0 96 0
30- E5 1 18
31- DF5 IU=1 IC=10 FAC=6.44e7

fatal error. detector no. 1 of tally 5 is not in any cell.

ring detector specifications
detector a0 r axis r0
1 0.00000E+00 9.60000E+01 z 0.00000E+00
 
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  • #2
Hi,
I am agree with mcnp :-)
Your geometry is very strange at the position x=0 y=96 z=0 it is a no man's land : nothing "le néant"
 
  • #3
It looks like the issue may be with your detector specifications. In the output file, it mentions that detector no. 1 of tally 5 is not in any cell. Looking at your input file, it seems like you have specified the material and geometry, but not the detector location. The "ring detector specifications" section shows that detector no. 1 should be located at a radius of 96cm along the z-axis. However, in your input file, you have not specified the location of this detector.

To fix this issue, you will need to add a detector card in your input file to specify the location of detector no. 1. It should look something like this:

32- F5Z:N 0 96 0
33- E5 1 18
34- DF5 IU=1 IC=10 FAC=6.44e7
35- SDEF POS=0 0 96 CEL=1 PAR=1

This will place the detector at a radius of 96cm along the z-axis, in the same cell as your Cf-252 source. Make sure to adjust the detector location and cell number accordingly to match your geometry.

Hope this helps!
 

Related to MCNP Why does it say F5 tally is not in any cell?

1. Why does MCNP say "F5 tally is not in any cell?"

MCNP may say this if you have not specified any cells for the tally to score in. The F5 tally is a surface tally and needs to be placed on a surface in order to record data.

2. How do I fix the "F5 tally is not in any cell" error in MCNP?

To fix this error, you need to specify which cells you want the F5 tally to score data in. This can be done by using the "F5" card before the tally card and specifying the cell numbers or names.

3. Can I use the F5 tally without specifying any cells?

No, the F5 tally needs to be placed on a surface and have cells specified in order to record data. Otherwise, it will not know where to score the data.

4. Why is the F5 tally important in MCNP?

The F5 tally is important because it allows you to record data on a surface, which can be useful in many applications. It also has the ability to score data in multiple directions, making it a versatile tool for analyzing radiation transport.

5. Are there any alternatives to using the F5 tally in MCNP?

Yes, there are other types of tallies in MCNP that can be used to score data, such as F4 and F8. These tallies have different scoring capabilities and may be more suitable for certain applications. It is important to understand the differences between them and choose the appropriate tally for your needs.

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