Hello everyone, I am currently working on creating an LFR core model (based on Westinghouse’s DLFR) for my bachelor’s thesis and I need little help with creating assembly geometry. I’ve managed to create an assembly fully filled with pins, but what I actually need is an assembly with a central beam tube (images attached), which could be used for hosting a finger absorption rod or instrumentation. In my case, there is not going to be any equipment inside the beam tube, so it’s just filled with lead.
I tried different ways to create this type of geometry but I can’t figure out how to do it, as it is my first project using OpenMC or any Monte Carlo code. Below is the code that I created. I would be very grateful if someone could point me to create this type of geometry.
Best,
Rafal
import matplotlib
import openmc
#-----------------------------------------------------------#
# 1) Materials definitions
#-----------------------------------------------------------#
# UO2 fuel at 17.5% wt enrichment
uo2_inner = openmc.Material(1, name='UO2_17.5')
uo2_inner.set_density('g/cm3', 10.97)
uo2_inner.add_nuclide('U235', 0.1526, 'wo')
uo2_inner.add_nuclide('U238', 0.7286, 'wo')
uo2_inner.add_nuclide('O16', 0.1188, 'wo')
# Helium for gap
helium = openmc.Material(3, name='Helium')
helium.set_density('g/cm3', 0.178)
helium.add_element('He', 1)
# Cladding
clad = openmc.Material(4, name='Cladding')
clad.set_density('g/cm3', 6.55)
clad.add_element('Zr', 1.0)
# Lead
lead = openmc.Material(5, name='Lead')
lead.set_density('g/cm3', 10.68)
lead.add_nuclide('Pb208', 0.524, 'wo')
lead.add_nuclide('Pb207', 0.221, 'wo')
lead.add_nuclide('Pb206', 0.241, 'wo')
lead.add_nuclide('Pb204', 0.014, 'wo')
materials = openmc.Materials([uo2_inner, helium, clad, lead])
materials.export_to_xml()
#-----------------------------------------------------------#
# 2) Geometry definitions - inner assembly
#-----------------------------------------------------------#
# Boundries and outer universe
all_lead_out=openmc.Cell(cell_id=200, fill=lead)
# Top & bottom of the assembly
core_height_z0 = openmc.ZPlane(surface_id=300, z0=-35, boundary_type='reflective')
core_height_z1 = openmc.ZPlane(surface_id=301, z0=35, boundary_type='reflective')
inner_assembly = openmc.model.hexagonal_prism(edge_length=17.09, orientation='x', boundary_type='transmission')
assembly_cell = openmc.Cell()
assembly_cell.region = inner_assembly & -core_height_z1 & +core_height_z0
# Create universes
all_lead_out_u=openmc.Universe(cells=[all_lead_out])
u_root = openmc.Universe()
# assembly cylinders
fuel_ir1 = openmc.ZCylinder(surface_id=410, r=0.3/2)
fuel_or1 = openmc.ZCylinder(surface_id=411, r=1.0/2)
clad_ir1 = openmc.ZCylinder(surface_id=412, r=1.03/2)
clad_or1 = openmc.ZCylinder(surface_id=413, r=1.16/2)
# assembly regions
central_hole_region1 = -fuel_ir1 & -core_height_z1 & +core_height_z0
fuel_region1 = +fuel_ir1 & -fuel_or1 & -core_height_z1 & +core_height_z0
gap_region1 = +fuel_or1 & -clad_ir1 & -core_height_z1 & +core_height_z0
clad_region1 = +clad_ir1 & -clad_or1 & -core_height_z1 & +core_height_z0
lead_region1 = +clad_or1 & -core_height_z1 & +core_height_z0
# assembly cells
central_hole_cell1 = openmc.Cell(cell_id=510, fill=helium, region=central_hole_region1)
fuel_cell1 = openmc.Cell(cell_id=511, fill=uo2_inner, region=fuel_region1)
gap_cell1 = openmc.Cell(cell_id=512, fill=helium, region=gap_region1)
clad_cell1 = openmc.Cell(cell_id=513, fill=clad, region=clad_region1)
lead_cell1 = openmc.Cell(cell_id=514, fill=lead, region=lead_region1)
u_inner = openmc.Universe(universe_id=3, cells=[central_hole_cell1, fuel_cell1, gap_cell1, clad_cell1, lead_cell1])
# Creating the hexagonal lattice
lat = openmc.HexLattice(name='assembly')
lat.center = (0., 0.)
lat.pitch = (1.36,)
lat.outer = all_lead_out_u
ring12 = [u_inner]*72
ring11 = [u_inner]*66
ring10 = [u_inner]*60
ring9 = [u_inner]*54
ring8 = [u_inner]*48
ring7 = [u_inner]*42
ring6 = [u_inner]*36
ring5 = [u_inner]*30
ring4 = [u_inner]*24
ring3 = [u_inner]*18
ring2 = [u_inner]*12
ring1 = [u_inner]*6
ring0 = [u_inner]*1
lat.universes = [ring12, ring11, ring10, ring9, ring8, ring7, ring6, ring5, ring4, ring3, ring2, ring1, ring0]
lat.orientation = 'x'
assembly_cell.fill = lat
out_assembly = openmc.Cell(fill=lead, region=~inner_assembly & -core_height_z1 & +core_height_z0)
#full_assembly = openmc.Universe(cells=[assembly_cell, out_assembly])
#-----------------------------------------------------------#
# 3) Geometry definitions - beam tube
#-----------------------------------------------------------#
beam_tube_prism = openmc.model.hexagonal_prism(edge_length=4.677, orientation='x', boundary_type='transmission')
beam_tube = openmc.ZCylinder(surface_id=414, r=7.6/2)
beam_tube_region1 = -beam_tube & -core_height_z1 & +core_height_z0
beam_tube_prism_region1 = beam_tube_prism & -core_height_z1 & +core_height_z0
beam_tube_cell1 = openmc.Cell(cell_id=515, fill=lead, region=beam_tube_region1)
beam_tube_prism_cell1 = openmc.Cell(cell_id=516, fill=clad, region=beam_tube_prism_region1)
out_beam_tube = openmc.Cell(fill=lead, region=~beam_tube_prism & -core_height_z1 & +core_height_z0)
beam_tube_u = openmc.Universe(cells=[beam_tube_cell1, beam_tube_prism_cell1, out_beam_tube])
u_root.add_cells([assembly_cell, out_assembly, beam_tube_cell1,beam_tube_prism_cell1, out_beam_tube])
geom = openmc.Geometry(u_root)
geom.export_to_xml()
plot2 = openmc.Plot()
plot2.filename = 'materials-xy'
plot2.basis = 'xy'
plot2.origin = [0, 0, 0]
plot2.width = [50,50]
plot2.pixels = [1000,1000]
plot2.color_by = 'material'
plots = openmc.Plots(plots=[plot2, ])
plots.export_to_xml()
openmc.plot_geometry(( plot2))
# OpenMC simulation parameters
batches = 20
inactive = 10
particles = 1000
settings = openmc.Settings()
settings.batches = batches
settings.inactive = inactive
settings.particles = particles
uniform_dist = openmc.stats.Box([-10,-10,-10],[10,10,10],only_fissionable=True)
settings.source = openmc.source.Source(space=uniform_dist)
settings.export_to_xml()
openmc.run()
