Why is My MCNP Program Outputting Too Small a Value?

  • #1
angfells
7
0
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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  • #2
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).
 
  • #3
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...
 

Related to Why is My MCNP Program Outputting Too Small a Value?

1. Why am I getting an unexpectedly small value in my MCNP output?

Small values in MCNP output could be due to several factors, including insufficient source particles, incorrect material definitions, or improper tally specifications. Ensure that your source definition is correct and that you have enough particles to achieve statistically significant results.

2. How can I increase the statistical accuracy of my MCNP simulation results?

To increase statistical accuracy, you can run more source particles, use variance reduction techniques, or increase the number of histories. Additionally, ensure that your tally regions are sufficiently large to capture enough interactions.

3. Could geometry errors lead to small values in my MCNP output?

Yes, geometry errors can lead to small values in the output. Ensure that your geometry is correctly defined and that there are no overlaps or gaps between cells that could cause particles to be lost or miscounted.

4. How do I verify that my material definitions are correct in MCNP?

Check your material definitions by comparing them with standard references or databases. Verify that the atomic densities, compositions, and cross-sections are correctly specified. Incorrect material definitions can significantly affect the simulation results.

5. What role does the tally specification play in the output values of an MCNP program?

The tally specification is crucial as it defines the quantities to be measured and the regions where measurements are taken. Ensure that the tally type, energy bins, and spatial regions are appropriately defined to capture the desired information accurately.

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