Problem encountered in drawing hexagonal regions: The boundary on the xy plane does not exist

User Support
1

Hello everyone, I have been conducting calculations for a hexagonal gas cooled reactor recently, but during the modeling process, openmc showed that I have no boundaries on the xy plane. I checked the code again but did not find any errors. Here is my code, please advise

import openmc
import os
from math import log10
import numpy as np
import math
import matplotlib.pyplot as pyplot

openmc.config['cross_sections'] ="/home/server/cs/nndc-b7.1-hdf5/cross_sections.xml"

fuel=openmc.Material(name='fuel')
fuel.add_element('U',1.0,enrichment=19.75)
fuel.add_nuclide('O16',2.0)
fuel.set_density('g/cc',10.3)
fuel.depletable=True

water=openmc.Material(name='water')
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16',1.0)
water.set_density('g/cc', 1.0)
water.add_s_alpha_beta('c_H_in_H2O')

co2=openmc.Material(name='co2')
co2.add_element('C',1.0)
co2.add_nuclide('O16',2.0)
co2.set_density('kg/m3',128.7)

Cr=openmc.Material(name='Cr')
Cr.add_element('Cr',1.0)
Cr.set_density('g/cm3',7.9)

Ni=openmc.Material(name='Ni')
Ni.add_element('Ni',1.0)
Ni.set_density('g/cm3',8.902)

Mo=openmc.Material(name='Mo')
Mo.add_element('Mo',1.0)
Mo.set_density('g/cm3',10.23)

Mn=openmc.Material(name='Mn')
Mn.add_element('Mn',1.0)
Mn.set_density('g/cm3',7.21)

Si=openmc.Material(name='Si')
Si.add_element('Si',1.0)
Si.set_density('g/cm3',2.33)

P=openmc.Material(name='P')
P.add_element('P',1.0)
P.set_density('g/cm3',1.82)

S=openmc.Material(name='S')
S.add_element('S',1.0)
S.set_density('g/cm3',2.069)

C=openmc.Material(name='C')
C.add_element('C',1.0)
C.set_density('g/cm3',2.281)

Fe=openmc.Material(name='Fe')
Fe.add_element('Fe',1.0)
Fe.set_density('g/cm3',7.86)


S316=openmc.Material.mix_materials([Cr,Ni,Mo,Mn,Si,P,S,C,Fe],[0.17,0.12,0.02,0.02,0.0075,0.00045,0.0003,0.0008,0.66095])

materials_file = openmc.Materials([fuel,water,co2,C,S316])
materials_file.export_to_xml()

# ------------------------------------------------------------------------------------------------------------------------------------------------
r_fuel=openmc.ZCylinder(r=0.425)
r_co2=openmc.ZCylinder(r=0.075)

z0=openmc.ZPlane(z0=0)
z1=openmc.ZPlane(z0=150)

fuel_cell=openmc.Cell(fill=fuel,region=-r_fuel & +z0 & -z1)
co2_cell=openmc.Cell(fill=co2,region=-r_co2& +z0 & -z1)
S316_cell=openmc.Cell(fill=S316,region=+r_fuel & +r_co2& +z0 & -z1)

fuel_universe=openmc.Universe(cells=(fuel_cell,S316_cell))

all_S316_cell=openmc.Cell(fill=S316)
outer_universe=openmc.Universe(cells=(all_S316_cell,))

lattice_fuel=openmc.HexLattice()
lattice_fuel.center=(0,0)
lattice_fuel.pitch=(1.2,)
lattice_fuel.outer=outer_universe

ring_1=[fuel_universe]*6
ring_in=[fuel_universe]

lattice_fuel.universes=[ring_1,
                        ring_in]
lattice_fuel.orientation='x'

component_region=openmc.model.HexagonalPrism(edge_length=1.2,orientation='x',origin=(0,0))
# component_region=-plan1 & +plan2 & -plan3 & +plan4 & -plan5 & +plan6& +z0 & -z1
component_cell=openmc.Cell(fill=lattice_fuel,region=-component_region&+z0&-z1)

u_root=openmc.Universe()
u_root.add_cells([component_cell])
geometry=openmc.Geometry(u_root)
geometry.export_to_xml()

plot=openmc.Plot.from_geometry(geometry)
plot.basis='xy'
plot.origin=(0,0,75)
plot.width=(2.4,2.4)
plot.color_by='material'
plot.to_ipython_image()
2

try this
plot=openmc.Plot
plot.basis=‘xy’
plot.origin=(0,0,75)
plot.width=(2.4,2.4)
plot.color_by=‘material’
plot.colors = colors = {
fuel: ‘red’,
S316: ‘green’,
co2:‘orange’,
}
openmc.plot_geometry()