Hello!
I am trying to run depletion calculations using materials in different values of temperatures. Those temperatures do not exist in the ENDF7.1 crosse section library. Therefore, I tried to use the Multipole method. However, I end up with the error below. It’s only working when I mix the Nearest or the Interpolation method with the Multipole. This makes me wonder, Is OpenMC using my set of temperatures or the temperatures in the ENDF7.1?
In the case that OpenMC is using the temperatures in the ENDF7.1, then how can I make it use my materials temperatures or any other temperature? Please.
My second question is related to the transfer of library from ace to h5 format using “openmc.data.IncidentNeutron.from_ace()”. It seems that when I transfer an ace file that contains multiple temperatures data, only the data related to the first temperature is converted to the h5 format. Therefore, I tried to use the method “add_temperature_from_ace()”. But, this required that the data for different temperatures should be placed in separate files, for example:
for suffix in [711, 712, 713, 714, 715, 716]:
u235.add_temperature_from_ace(‘92235.{}nc’.format(suffix))
Then, would someone know how to convert all the ace data at once?
