How to check the isotropy of the source in MCNP?

  • Thread starter Thread starter angfells
  • Start date Start date
  • Tags Tags
    Mcnp Mcnp5
AI Thread Summary
To verify the isotropy of a point source, one effective method involves surrounding the source with a transparent sphere and segmenting it into equal areas, similar to an orange. This allows for accurate tallies of neutron flux across different surfaces. The goal is to ensure that the particle flow is evenly distributed throughout the source volume. The discussion emphasizes the importance of rigorous checks to confirm isotropic behavior. Implementing this approach should provide a clear assessment of isotropy.
angfells
Messages
7
Reaction score
0
Hello everyone!
I need to make sure that my source is isotropic. How can I check that?
I have point source pos -11 0 0 erg=d1 with Maxwellian spectrum of energy and some surfaces through which neutron flux passes.
 
Engineering news on Phys.org
How rigorous does the check need to be?

You could make everything else transparent, put a sphere round your source, segment it like an orange and then section the orange in half. Every surface would then have equal area and you could do tallies.
 
Alex A said:
How rigorous does the check need to be?

You could make everything else transparent, put a sphere round your source, segment it like an orange and then section the orange in half. Every surface would then have equal area and you could do tallies.
To really see that the source is isotropic, i.e. the particle flow is equally distributed throughout the source volume. Everything is already transparent. Thanks for your answer, I'll try it.
 
Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information(power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.) for Material 2 is zero, while Material 1...
Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6? If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning. I'd be really...

Similar threads

Replies
4
Views
980
Replies
4
Views
2K
Replies
2
Views
3K
Replies
17
Views
3K
Replies
1
Views
2K
Replies
1
Views
2K
Back
Top