Dear experts, I want to find fission products in OpenMC by changing the fuel pins in Lattice, for example by simultaneously depletion 78 pins of different fuel and 234 pins of different fuel pins for the VVER reactor. How can I make adjustments for this in OpenMC?How should we add the total fuel volume to the input?
fuel1 = openmc.Material(material_id=1, name='Fuel1')
fuel1.add_nuclide('U234', 0.00000497749532321889)
fuel1.add_nuclide('U235', 0.00117242970759577)
fuel1.add_nuclide('U238', 0.0232992046466795)
fuel1.add_nuclide('O16', 0.048953223699197)
fuel1.volume = 234 * pi * fuel_radius**2
fuel2 = openmc.Material(material_id=6, name='Fuel2')
fuel2.add_nuclide('Th232', 0.0114040986857742)
fuel2.add_nuclide('O16', 0.0483170246105698)
fuel2.add_nuclide('Pu239', 0.00861126468135872)
fuel2.add_nuclide('Pu240', 0.00268315477573431)
fuel2.add_nuclide('Pu241', 0.000969806702233638)
fuel2.add_nuclide('Pu242', 0.000381147686066511)
fuel2.volume = 78 * pi * fuel_radius**2