Please help with this MCNP program -- in the output response I get too small a value

[angfells]

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. My code and output files are below.

 

[Alex A]

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).

 

[angfells]

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...