I am very interested in using OpenMC as a design tool for a critical benchmark in verifying neutron cross sections. For this purpose, sensitivity and uncertainty analysis capability will be immensely useful in verifying how keff will react to slight changes in cross section, dimensions and composition of the assembly, and optimize for this sensitivity. Thus far, I have built an OpenMC model of the critical assembly and verified keff numbers against MCNP, and the results agrees well, but I will not be able to move on to the optimization phase without S&U analysis.
As you may know, such a functionality already exists within MCNP and the SCALE code package. Upon doing some literature research, I came across the paper titled “Development of continuous-energy sensitivity analysis capability in OpenMC” published on the Annuals of Nuclear Energy. It outlines the implementation of the CLUTCH-IFP method in OpenMC and even provides some validation against critical benchmarks very similar to the one I am designing. In my excitement, I contacted the authors but have not been able to obtain a copy of the OpenMC code they used. To my knowledge, there has not been a published version of OpenMC with this method implemented. May I ask if this is a feature in active development? I will appreciate any advice to point me in the right direction, thanks!
- Mason Yu