Below please see the material specification for a molten salt reactor fuel. I am using the
materials.temperature = 1200.0
line to change the analyzed temperature, but no change occurs in the calculated keff. As you can see from the below snippet, I am using the 80x library.
What am I doing incorrectly?
Thank you, and I apologize if this is a duplicate question.
Brian
From Experience with the Molten-Salt Reactor Experiment - Haubenreich and Engel
LiF-BeF2-ZrF4-UF4 (65 29.1 5.0 0.9 mole %, U235 33% enrichment
FuelSalt = openmc.Material(1,“FuelSalt”)
FuelSalt.add_nuclide(‘U235’,0.003,‘ao’)
FuelSalt.add_nuclide(‘U238’,0.006,‘ao’)
FuelSalt.add_nuclide(‘Li7’, 0.69,‘ao’)
FuelSalt.add_nuclide(‘F19’, (.691 + .2912 + .054 + .014),‘ao’)
FuelSalt.add_nuclide(‘Be9’, .291,‘ao’)
FuelSalt.add_nuclide(‘Zr91’,.05,‘ao’)
FuelSalt.set_density(‘g/cm3’, 2.3)
#Graphite Moderator
Graphite = openmc.Material(2,“Graphite”)
Graphite.add_nuclide(‘C12’,1.0,‘ao’)
Graphite.set_density(‘g/cm3’, 2.1)
#Hastelloy N alloy - specified as weight percent in Haynes documentation
Hastelloy = openmc.Material(3,‘Hastelloy’)
Hastelloy.add_nuclide(‘Ni59’,0.71,‘wo’)
Hastelloy.add_nuclide(‘Mo96’,0.16,‘wo’)
Hastelloy.add_nuclide(‘Cr52’,0.07,‘wo’)
Hastelloy.add_nuclide(‘Fe56’,0.04,‘wo’)
Hastelloy.add_nuclide(‘C12’ ,0.0006,‘wo’)
Hastelloy.add_nuclide(‘Si28’,0.01,‘wo’)
Hastelloy.add_nuclide(‘Mn55’,0.008,‘wo’)
Hastelloy.add_nuclide(‘V51’, 0.0014,‘wo’)
Hastelloy.set_density(‘g/cm3’, 8.86)
materials=openmc.Materials([FuelSalt,Graphite,Hastelloy])
##Changing this temperature specification does not change the keff???
materials.temperature = 1200.0
materials.cross_sections="/home/brian/OpenMC_data/NuclearData/lib80x_hdf5/cross_sections.xml"
materials.export_to_xml()