Where Are the Other Detectors for F5 Tally in MCNP?

In summary, the code allows for multiple detectors to be assigned to a single F5 tally, but the output file only lists the tally result for the first detector.
  • #1
Will_007
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1
TL;DR Summary
where is the output for multiple F5 tally detectors
The MCNP manual states that you can have multiple detectors for a single F5 tally. Say you have f15:n x1 y1 z1 r x2 y2 z2 r.....Thing is, my output file only lists the tally result for the first f5 detector (x1,y1,z1). Where are other detectors for this tally? Is there a reason code developers enabled this method of defining multiple detectors vs just using f15, f25, f35....for each one?

Thanks,
Will
 
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  • #2
f15, f25, f35 should all work and give separate answers, I wonder if you specify multiple rings or points on the same tally if it sums all the calculations to a single answer. I might do a test. I'd say the code is not 'clever' but internally it's extremely clever. What it's not is user friendly. It's derived from an old code and it's still in FORTRAN, which has limited string handling capabilities, everything has to be quite explicit; syntax is fairly brittle and error checking is woefully lacking.
 
  • #3
Hey Alex - so yes, F15, F25....can be used separately but this option allows you to use F15 and then list multiple detector locations (not ring or a sum...the results appear to be as if you separately assigned a tally number to each). Discussion in the manual is limited, but in latest manual online (MCNP 6.3), this is the statement below. It looks like only the first detector belonging to the tally is in the output file, the rest are in the .m file but minimal labels (not user friendly at all)
1675206546484.png
 
  • #4
I have almost no experience with MCNP6, so I've done some reading. There also isn't much in the way of a glossary. I understand a tally is an F card, a bin is when it's split into multiple results by energy, position or time and a detector is an object that has to be inserted in addition to the existing geometry and cell cards.

F1,2,3,4,6 and 7 don't involve detectors, they use an existing object. MCNP5 and later allows many objects on the same tally and how the results are shown depends upon parentheses.

MCNP5 only allows one ring or point per tally but the same note is in the MCNP5 manual, just meaning F5 can do point arrays (and maybe other reasons) so don't expect to do 20 (the detector limit for that version) F5 tallies under all circumstances.

X 2.6.0, 6.2 and 6.3 manuals allow multiple ring and point detectors per tally and the expected behavior is that is write a tally total as well as individual results for each detector unless this is suppressed with the 'ND' keyword.

The 6.1 and 6.1.1 manuals are maybe unhelpful, and they appear to be derived from the MCNP5 manual, so the behavior might not match the code.

So yeah, unless you are using the 'ND' keyword you should see a result for every detector on the line according to the manual. I also have no idea what the .m file is. :)
 
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  • #5
thanks - .m file is the MCTAL file (obviously! :) )...one day when LANL makes this a user-friendly code it will transform it from formula one to bentley
 
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FAQ: Where Are the Other Detectors for F5 Tally in MCNP?

Where can I find the other detectors for F5 tally in MCNP?

The other detectors for F5 tally in MCNP can be specified within the input file. You need to define additional F5 tally cards with different detector coordinates or parameters. The MCNP manual provides detailed instructions on how to set up multiple detectors.

How do I specify multiple F5 tallies in MCNP?

To specify multiple F5 tallies, you need to use separate F5 tally cards for each detector. Each F5 tally card should have a unique tally number and the coordinates or parameters specific to the detector location. Ensure that each tally card is properly formatted according to the MCNP input requirements.

Can I use the same F5 tally for different energy ranges?

Yes, you can use the same F5 tally for different energy ranges by specifying energy bins within the tally card. You can define energy bins using the E card associated with the F5 tally. This allows you to obtain tally results for different energy ranges using the same detector.

What is the purpose of using multiple F5 tallies?

Using multiple F5 tallies allows you to measure the flux or response at different locations or under different conditions within your simulation. This can provide a more comprehensive understanding of the radiation field and help optimize the design or analysis of your system.

How do I analyze the results from multiple F5 tallies?

After running the MCNP simulation, the results from multiple F5 tallies will be available in the output file. Each tally will have its own section in the output, identified by its unique tally number. You can analyze these results by comparing the flux or response values at different detector locations or conditions as specified in your input file.

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