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thinking about how to use this code with openmc and I often have the situation where the fusion reactor is quotes as 1GW fusion power and then there is a need to find number of fusion reactions from t…
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## Description of issue / requirement to address
The STP CAD output files need material metadata attached to volumes or groups of volumes so that it can be more easily used in neutronics analysis. Th…
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![image](https://github.com/user-attachments/assets/9386d1b7-f1b0-4e33-a824-62ad3241c035)
## Working groups:
#### Plasma Core (inside the LCFS)
#### Power Handling and Electromagnetic Loa…
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Add tests for the radiation_transport/neutronics folder as there are none
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I just came across the repo Numjuggler: https://github.com/inr-kit/numjuggler?tab=readme-ov-file. We should probably mention them as well for a possible alternative in our JOSS paper submission.
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Please note that I'm not too knowledgeable with the code, so feel free to comment angry messages and close the issue if this is in error :)
For example, for ENDF-B-VII.1-neutrons n-008-O_017.endf,…
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## Description of issue / requirement to address
Add more tests to Radiation model its not particularly well covered at the moment
- [ ] #3328
- [x] #3329
- [ ] #3330
- [x] #3331
- [x] #…
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OpenMC is the default code for the DAGMC export function but it can be useful to support other codes.
DAGMC geometry tags are slightly different
in openmc we can use material strings, these stri…
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PRs #766 and #767 both experience what seems to be randomly failing tests.
See : https://github.com/ukaea/paramak/pull/766/checks?check_run_id=2048177525
This is the error message.
```
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