Hi,
The picture on the left is the result I got from pin depletion in Example, I hope to get the picture on the right (the initial value of KEFF is very high), but there is no obvious difference in the result after I modified the U235 content, may I ask what I can do?
One issue that you’d probably want to fix is that your timesteps are very long. The stability of the numerical methods used for depletion can be sensitive to the length of the timesteps. If they are too long, the answers will not be reliable.
In general, the keff value at t=0 is determined by your initial material compositions and overall geometry arrangement. If you want a higher keff at t=0, you need to change either the materials or the geometry.
Thank you for your reply. Yes! My problem is how to change the material and geometry to increase the initial keff. I tried to change the ratio of U235 and the radius of the fuel, but they didn’t seem to work.
Here’s the material and geometry section of my code:
fuel = openmc.Material(name=“uo2”)
fuel.add_nuclide(‘U238’,0.95)
fuel.add_nuclide(‘U235’,0.05)
fuel.add_nuclide(‘O16’,2.0)
fuel.set_density(‘g/cm3’,12)
clad = openmc.Material(name=“clad”)
clad.add_element(“Zr”, 1)
clad.set_density(“g/cc”, 6)
water = openmc.Material(name=“water”)
water.add_element(“O”, 1)
water.add_element(“H”, 2)
water.set_density(“g/cc”, 1.0)
water.add_s_alpha_beta(“c_H_in_H2O”)
materials = openmc.Materials([fuel, clad, water])
radii = [0.42, 0.45]
pin_surfaces = [openmc.ZCylinder(r=r) for r in radii]
pin_univ = openmc.model.pin(pin_surfaces, materials)
bound_box = openmc.rectangular_prism(0.62, 0.62, boundary_type=“reflective”)
root_cell = openmc.Cell(fill=pin_univ, region=bound_box)
root_univ = openmc.Universe(cells=[root_cell])
I think you did not use the correct parameter in the rectangular_prism() function.
Why not try this:
bound_box = openmc.rectangular_prism(1.24, 1.24, boundary_type=“reflective”)
I also recommend you use plot capability to debug modeling error.