We are looking to see if we can use OpenMC for irradiation experiment analysis for ATR and TREAT reactors at INL. I was wondering if someone could quickly answer what requirements OpenMC would be able to satisfy and what could not.
|F.1|The software shall provide depletion/activation of isotopes specified by the user for the following reactions: (n,fission), (n,γ), (n,alpha), (n,p), and (n,xn’).|
|F.2|The software shall track isotopic depletion and transmutation of materials in user specified regions.|
|F.3|The software shall be able to accurately calculate an eigenvalue (keff) for the geometric configuration for the ATR, ATRC, and TREAT.|
|F.4|The software shall allow for the reactor operating cycle simulation and changes in conditions such as reactor power, shutdown, rotating movement of control shims and axial movement of control shims.|
|F.5|The software shall be able to calculate the energy deposition from prompt and delayed neutrons and photons as the result of fission in various user specified regions within the reactor.|
|F.6|The software shall calculate fission density in a fueled region of interest|
|F.7|The software shall allow for the user to input compositions for irradiated fuel elements and experiments.|
|F.8|The software shall be able to calculate the neutron flux for multiple user specified energy bins for a geometric configuration similar to the ATR.|
|F.9|The software shall provide results for eigenvalue and neutron/photon heating rate tallies with statistical uncertainty. |
|F.10|The software shall allow the development of ATR, ATRC, and TREAT to support efficient analysis. |
|F.11|The software shall calculate neutron kinetics parameters.|
|F.12|The software shall Doppler broaden neutron cross sections.|