I was trying to OpenMC to calculate for the BN-600 MOX Core Benchmark (PHASE4).However, there was a problem. In its material, there is a nuclide of 239FP, which means fission product of 239Pu.I am curious to know how I would handle this nuclide in OpenMC?Also, any other advice would be highly appreciated. Thank You!
OpenMC doesn’t have pre-made lumped fission product cross sections. You will have to take an approach similar to what is done by the other benchmark participants: identify a suitable approximation that meets the accuracy needs of your application.
Some options include:
Select a single fission product nuclide to act as a surrogate for all the fission products. Since BN-600 is a fast reactor, you may be able to get “close enough” results.
Take the Pu239 yield curve from some reference and use those isotopes and their relative concentrations. This is better than option 1, but still an approximation. One of the benchmark participants in the pdf you linked umstated they used this approach.
Run explicit depletion calculations (likely in a smaller model) to find a more accurate distribution of fission products and use those instead. This is closer to what is done in production.
Hope this helps, and sorry it’s not a straight forward answer.
Thanks for you reply. Your suggestions are very helpful to me. I would use the option 2 for time and computing resources.
But I have one more question. If I get the Pu239 yield curve through the fast reactor burnup chain (chain_endfb71_sfr.xml) provided by OpenMC, Is it appropriate? Is there anything that needs special attention?
Thanks again for your help.
@TYJ The chain_endfb71_sfr.xml has fission product yields listed for multiple energies, so you’ll need to make sure that you select the yields for the appropriate energy. The only “sfr” specific information in the chain file is the capture branching fractions, which were computing in a fast spectrum.