Why is My MCNP Program Outputting Too Small a Value?

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SUMMARY

The discussion centers on an issue with the MCNP (Monte Carlo N-Particle Transport Code) program where the user is receiving unexpectedly low output values. The user is simulating neutron flux with a surface source and water moderator, but the results are per source particle and indicate that neutrons are traveling through empty space. The key takeaway is that the output reflects the number of neutrons passing through specified surfaces, which in this case is correctly calculated as 1.0 for all neutrons emitted from surface 15 passing through surface 3.

PREREQUISITES
  • Understanding of MCNP (Monte Carlo N-Particle Transport Code) version 6 or later
  • Knowledge of neutron transport theory
  • Familiarity with tally results and their interpretation in MCNP
  • Basic concepts of material interactions in radiation transport
NEXT STEPS
  • Review MCNP documentation on tally results and their significance
  • Learn about neutron flux calculations in MCNP
  • Explore the effects of different moderators, including vacuum, on neutron transport
  • Investigate common pitfalls in MCNP simulations and how to avoid them
USEFUL FOR

This discussion is beneficial for physicists, nuclear engineers, and researchers using MCNP for neutron transport simulations, particularly those troubleshooting output discrepancies.

angfells
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Hello everyone!
I have some troubles with my MCNP programm:
I have a source, a moderator and a tally. The source is surface, the moderator is water (but I need to calculate for vacuum as well). Only neutrons are used in this task. The neutron flux is unidirectional. I take 1e6 the number of stories, but in the output response I get too small a value. I've tried everything, I don't understand why it's happening:cry:. My code and output files are below.
 

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Hi, welcome to physicsforums.

All tally results are per source particle. You are not using a material in the problem so all the neutrons are traveling through empty space. So the current question is asking how many neutrons made on surface 15 pass through surface 3 and the answer is 1.0 (all of them).
 
Alex A said:
Hi, welcome to physicsforums.

All tally results are per source particle. You are not using a material in the problem so all the neutrons are traveling through empty space. So the current question is asking how many neutrons made on surface 15 pass through surface 3 and the answer is 1.0 (all of them).
Thanks a lot! Probably I misinterpreted the output file...
 

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