Why is My MCNP Program Outputting Too Small a Value?

In summary, the issue of an MCNP program outputting too small a value can stem from various factors, including incorrect input parameters, such as geometry, materials, or source definitions. It may also arise from insufficient statistics due to a low number of histories or an inappropriate tally setup. Additionally, issues related to normalization or scaling factors can contribute to unexpectedly low results. To resolve this, users should carefully review their input files, ensure adequate particle histories, and verify tally configurations to obtain accurate output values.
  • #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.
 

Attachments

  • test1 — копия (2).txt
    1.1 KB · Views: 80
  • outv — копия.txt
    38.2 KB · Views: 81
Engineering news on Phys.org
  • #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...
 

FAQ: Why is My MCNP Program Outputting Too Small a Value?

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.

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.

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.

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.

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.

Similar threads

If MCNP tallies are normalized, shouldn't they be <1?
Replies
5
Views
2K
What is a unit of time listed in MCNP6 PTRAC output file?
Replies
1
Views
2K
Troubleshooting MCNP6: 'Bad Trouble in Subroutine Source' & More
Replies
2
Views
3K
An extremely basic question on MCNP
Replies
2
Views
2K
MCNP6: error -- "bad trouble in imcn" is usually tangled with spacing error?
Replies
6
Views
3K
MCNP and simple nuclear physics
Replies
9
Views
2K
MCNP Modelling Detector Energy Response Vs Empirical Results
Replies
3
Views
639
"No tally, no plot" error MCNP6
Replies
24
Views
5K
Why is my SSW/SSR code not producing readable output from a KCODE run?
Replies
3
Views
2K
[MCNP] Lost too much Keff with Burnup card
Replies
5
Views
2K

Hot Threads

Recent Insights

Back
Top