Hi all,
I am modelling a pin-cell similar to the one here:
Pin cell plot, Serpent
The problem I have is a very big discrepancy between the OpenMC model and a Serpent model.
The Serpent model has the following geometry:
Geometry Built in Serpent
pin unit
gap 0.100
fuel 0.450
gap 0.465
clad 0.525
lead% Pin pitch is 1.35975
surf 1 hexyc 0.0 0.0 0.679875
surf 2 sqc 0.0 0.0 1.5%%%%%%%%%%%%%%%% % ------- Cells %%%%%%%%%%%%%%%%cell 1 0 fill unit -1
cell 2 0 gap 1 -2
cell 3 0 outside 2 % outside world
Now I am new to OpenMC. I modelled this geometry like so in OpenMC:
Geometry Built in OpenMC
# Create cylinders for the gap, fuel, and clad
gap_radius = openmc.ZCylinder(x0=0.0, y0=0.0, r=0.100)
fuel_radius = openmc.ZCylinder(x0=0.0, y0=0.0, r=0.450)
gap2_radius = openmc.ZCylinder(x0=0.0, y0=0.0, r=0.465)
clad_radius = openmc.ZCylinder(x0=0.0, y0=0.0, r=0.525)
# Create a hexagonal to surround the pin
hexagonal = openmc.model.HexagonalPrism(0.679875, orientation='x')
# Create box to surround the geometry
box = openmc.model.RectangularPrism(3.0, 3.0, boundary_type='reflective')
# Create a Universe to encapsulate a pin-cell
pin_cell_universe = openmc.Universe(name='pin_cell')
# Create different Cells and add them to the universe
#gap
gap_cell = openmc.Cell(name='GAP')
gap_cell.fill = gap
gap_cell.region = -gap_radius
pin_cell_universe.add_cell(gap_cell)
#fuel
fuel_cell = openmc.Cell(name='FUEL')
fuel_cell.fill = fuel
fuel_cell.region = -fuel_radius & +gap_radius
pin_cell_universe.add_cell(fuel_cell)
#gap2
gap2_cell = openmc.Cell(name='GAP2')
gap2_cell.fill = gap
gap2_cell.region = -gap2_radius & +fuel_radius
pin_cell_universe.add_cell(gap2_cell)
#clad
clad_cell = openmc.Cell(name='CLAD')
clad_cell.fill = clad
clad_cell.region = -clad_radius & +gap2_radius
pin_cell_universe.add_cell(clad_cell)
#lead
lead_cell = openmc.Cell(name='LEAD')
lead_cell.fill = lead
lead_cell.region = -hexagonal & +clad_radius
pin_cell_universe.add_cell(lead_cell)
#gap3
gap3_cell = openmc.Cell(name='GAP3')
gap3_cell.fill = gap
gap3_cell.region = +hexagonal & -box
pin_cell_universe.add_cell(gap3_cell)
# Create Geometry and set root Universe
model.geometry = openmc.Geometry(pin_cell_universe)
And noting that one can have better ways of doing it in OpenMC of course (or maybe even correct, I am not sure if this is correct).
Here is the geometry part of the model file. I am not sure about the hexagonal prism individual 6 planes are in the correct locations? seems like extensive use of sqrt(3) is being applied. This could be the problem I guess?
Model.xml geometry settings
<geometry>
<cell id="1" material="4" name="GAP" region="-1" universe="1"/>
<cell id="2" material="3" name="FUEL" region="-2 1" universe="1"/>
<cell id="3" material="4" name="GAP2" region="-3 2" universe="1"/>
<cell id="4" material="5" name="CLAD" region="-4 3" universe="1"/>
<cell id="5" material="6" name="LEAD" region="-5 6 -7 -10 8 9 4" universe="1"/>
<cell id="6" material="4" name="GAP3" region="~(-5 6 -7 -10 8 9) (11 -12 13 -14)" universe="1"/>
<surface coeffs="0.0 0.0 0.1" id="1" type="z-cylinder"/>
<surface coeffs="0.0 0.0 0.45" id="2" type="z-cylinder"/>
<surface coeffs="0.0 0.0 0.465" id="3" type="z-cylinder"/>
<surface coeffs="0.0 0.0 0.525" id="4" type="z-cylinder"/>
<surface coeffs="0.5887890213979452" id="5" type="y-plane"/>
<surface coeffs="-0.5887890213979452" id="6" type="y-plane"/>
<surface coeffs="1.7320508075688772 1.0 0.0 1.1775780427958904" id="7" type="plane"/>
<surface coeffs="-1.7320508075688772 1.0 0.0 -1.1775780427958904" id="8" type="plane"/>
<surface coeffs="1.7320508075688772 1.0 0.0 -1.1775780427958904" id="9" type="plane"/>
<surface coeffs="-1.7320508075688772 1.0 0.0 1.1775780427958904" id="10" type="plane"/>
<surface boundary="reflective" coeffs="-1.5" id="11" name="minimum x" type="x-plane"/>
<surface boundary="reflective" coeffs="1.5" id="12" name="maximum x" type="x-plane"/>
<surface boundary="reflective" coeffs="-1.5" id="13" name="minimum y" type="y-plane"/>
<surface boundary="reflective" coeffs="1.5" id="14" name="maximum y" type="y-plane"/>
</geometry>
I plotted the configuration for verification, and here is what I get using:
plot = openmc.Plot()
plot.color_by = 'material'
plot.basis = 'xy'
plot.origin = (0.0, 0.0, 0.0)
plot.width = (3., 3.)
plot.pixels = (400, 400)
plot.colors = {
gap: 'grey',
fuel: 'red',
clad: 'black',
lead: 'blue'
}
The plot will be displayed in the next reply as discord does not allow new users to embed more than one media item in a post.
The discrepancy is very big:
Serpent: 1.35712 +/- 0.00020
OpenMC: 1.41382 +/- 0.00135
-
I verified the materials between the two codes many times, so I think it is the geometry and how I am doing it in OpenMC.
-
Inspecting the Serpent geometry, the
sqcis 1.5 which is 1/2 the width of the square prism. I used 3 in the OpenMC geometry. -
Same for the hexagonal prism, Serpent accepts half the pitch (or the long side), while OpenMC accepts the hexagonal side (which is 1/2 the long side or the pitch for equilateral).
-
Orientation is also different between Serpent and OpenMC, I think I took this into account as well (cf. Figures).
-
I also tried using transmission BC on the hexagonal prism in OpenMC, but I think it is already applied by default (and does not make sense of course, but I was just trying).
Thank you,
