Why is My MCNP Program Outputting Too Small a Value?

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The user is experiencing issues with their MCNP program, specifically receiving unexpectedly low output values despite running 1 million histories. The setup includes a surface source, water moderator, and a tally, with a focus on neutron flux in a vacuum scenario. It is clarified that tally results are per source particle, and since no material is involved, all neutrons are traveling through empty space. The user acknowledges a possible misinterpretation of the output file. Understanding the tally results in relation to the source particle is crucial for accurate analysis.
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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