Help debugging a geometry-related error in my MCNP input deck

AI Thread Summary
The discussion revolves around troubleshooting a geometry-related error in an MCNP input deck, likely caused by surface overlaps in a complex lattice of spheres. The user is attempting to model a TRISO fuel compact using body-centered cubic (BCC) lattice cells, focusing on a cylindrical geometry. Suggestions include simplifying the lattice representation and considering homogenization for easier modeling and reduced computational time. The user ultimately decides to pursue a different method for their project. The thread concludes with encouragement to share any new findings.
MadGander
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TL;DR Summary
MCNP geometry error
I'm looking for someone to help troubleshoot my MCNP input deck. I'm getting a geometry related error most likely due to some sort of surface overlap. Haven't been able to identify the issue myself, so I'd appreciate a secondary check. Thanks!
 

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The input file is incomplete and is crashing for me. It seems to be a big lattice full of spheres. It's very verbose, with a lot of copy transpose.

I would suggest writing the lattice as a lattice. If you describe it I should be able to help.
 
Alex A said:
The input file is incomplete and is crashing for me. It seems to be a big lattice full of spheres. It's very verbose, with a lot of copy transpose.

I would suggest writing the lattice as a lattice. If you describe it I should be able to help.
Im trying to model a TRISO fuel compact using BCC lattice cells. I'm focusing specifically on the x-y plane to produce a cylindrical geometry. Depending on exact location, the periphery cells have some sort of pseudo-BCC structure to form the outer edge of the compact. Because of this, the periphery cells have certain void corners where no 1/8th particle slice exists. Appreciate the assistance.
 
If this is, for example, a 17x17 lattice then we can model this. The spheres are tiny though, under 1mm and I can't imagine this amount and geometry reaching criticality. If this is a set problem and you have to solve it this way we can do it. If what you want is the answer from MCNP on a running reactor with this material as fuel I would suggest you homogenise. If you have a structure that is 10cm across, neutrons will not care much about the fine structure, only the quantities. Working out quantities of elements in bigger units and building the reactor from the big units is easier to model and needs less computer time.

So what are all the relevant parameters of the problem, what answers do you need and how do you want to solve it?
 
Thread can be closed. I ended up ditching this route and am pursuing a different method.
 
If it's something you can share, please do consider sharing it. All the best with the new method!
 
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...

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