Hello everyone,
I wanted to ask if there’s any information available about using OpenMC results and deploying it using FISPACT or vice versa, specially in fusion engineering analysis. Thank you so much for your help!
I did quite a few comparisons when I worked at CCFE and had access to FISPACT. The
FISPACT Python API created by Tom Stainer made the process of inputting and extracting results super easy using Python. One just need a energy spectrum, material and irradiation schedule and nuclear data. I would take a look at the api_manual.pdf included in version 5.0 FISPACT-II 5.0, Inventory Simulation Platform for Nuclear Observables and Materials Science
These days I use the depletion inbuilt into OpenMC as it does all the fusion inventory analysis I need.
However I would be super interested to hear what simulations you would want to do with FISPACT that the inbuilt depletion solver can’t do just in case I’ve missed a feature.
There are a few UKAEA/ CCFE people on this forum so they might pitch in, but you can also ask on the FISPACT forum FISPACT-II Forum - Index page
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Thank you so much for your help and insight! This is very helpful! I was wondering if it’s possible to use FISPACT as a part of simulation of fast neutron irradiators based on fusion device, to obtain the strength and spatial distribution of the neutron source. Can I do that with only OpenMC?
FISPACT handles the activation/transmutation part of the calculation (that is, solving the Bateman equations) and will not provide you any information on a neutron source. The strength/spatial distribution of the neutron source is required as a user input when running OpenMC. For fusion applications (tokamaks specifically), the neutron source is informed by plasma physics calculations and if my understanding is correct, the source is often parameterized based on the characteristics of the plasma. Perhaps @Shimwell or another fusion expert can elaborate.
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I agree with you @paulromano
Also decay neutron sources (if that is what we are talking about) are quite rare in fusion. Hybrid blankets with spontaneous fission from fissile material can emit neutrons or beryllium which can generate neutrons from high energy decay gammas are the only cases I can think of in my sleep deprivation state.
In general we use FISPACT to find the quantity of isotopes produced during irradiation and the resulting particle spectrum (normally gamma but other particles are also included) from the activated material. These are both things that OpenMC can perform.
However it is good to compare results with other codes. However for actual usability the deployability, permissive licence and scalability of OpenMC make it quicker to get up and running with. The tight integration of the Bateman solver in OpenMC with the Monte Carlo code mean that a pure OpenMC solution will likely have a performance advantage for this type of simulation.
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Hi - sorry to revive an old discussion, but I’ve been tasked with investigating ‘OpenMC vs FISPACT’ for fusion blanket activation purposes. I was wondering if you could give some advice - I’ve looked around a bunch of sources but haven’t seen any direct comparisons, and my understanding of the codes is relatively limited.
What are the main limitations of OpenMC depletion applied to a blanket - I’ve read here that openmc is relatively immature for some fusion purposes, but I know much work has been done on this front. I expect the nuclear data may play a large role? And what (if any) differences in results would you expect to see between OpenMC and FISPACT simulations? Thanks 
@srichr221 I’ve been doing OpenMC vs FISPACT comparisons for a paper that is going to be submitted shortly along with @eepeterson. In particular, we’ve looked at predicted material compositions after irradiation/decay as well as the resulting gamma spectra. The short answer is that given equivalent nuclear data, the two codes will produce predictions that are very close to one another. There are a few subtle differences — for example, FISPACT uses an approximate gamma spectrum when no discrete lines are available in the underlying ENDF evaluation, whereas OpenMC will try to use a continuous spectrum from the ENDF evaluation if available.
Another subtle difference is that generally you provide FISPACT with multigroup fluxes, and it has to make some assumption about the within-group flux in order to calculation reaction rates. With OpenMC, it’s easy to generate the exact microscopic cross sections that are used to calculate reaction rates. @eepeterson may have more to say on the impact of that.
Keep in mind that FISPACT has quite a few features that are not yet replicated in OpenMC (e.g., pathway and uncertainty analysis), so depending on your needs OpenMC may or may not be a suitable alternative. If there is a feature you really need that is not present in OpenMC though, do let us know!
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