Hello all,
I think there is something funny going on with the plotting of openmc. below is part of the code I am using for plotting the mesh which I am trying to use to tally fission. when I use this the mesh is shown in the figure is obtained which I don’t think is correct.
also if there are suggestions on how to tally for assembly power other than the method I am using, please do suggest.
thank you
import openmc
import constants as c
import assemblies as a
import pincells as p
#import os
#os.environ[‘OPENMC_CROSS_SECTIONS’]=‘/mnt/d/Work/XS/endfb-vii.1-hdf5/cross_sections.xml’
A0 = a.a0(); B0 = a.b0(); C0 = a.c0(); Z0 = p.inf_water(); V0= p.inf_void()
B1 = a.b1(); C1 = a.c1()
B2 = a.b2(); C2 = a.c2()
B3 = a.b3(); C3 = a.c3()
core= openmc.RectLattice(name=‘quarter core’)
core.pitch = (c.latticePitch, c.latticePitch)
core.lower_left = (-c.latticePitch/2, -c.latticePitch*19/2)
core.universes = [ [A0,A0,C3,A0,B1,A0,B3,C2,B0,Z0],
[A0,B3,A0,B3,A0,B1,A0,B3,C0,Z0],
[C3,A0,C2,A0,C3,A0,C3,B1,B0,Z0],
[A0,B3,A0,B3,A0,B3,A0,B2,C0,Z0],
[B1,A0,C3,A0,C2,A0,B1,C0,Z0,V0],
[A0,B1,A0,B3,A0,B3,C1,C0,Z0,V0],
[B3,A0,C3,A0,B1,C1,C0,Z0,V0,V0],
[C2,B3,B1,B2,C0,C0,Z0,V0,V0,V0],
[B0,C0,B0,C0,Z0,Z0,V0,V0,V0,V0],
[Z0,Z0,Z0,Z0,V0,V0,V0,V0,V0,V0]]
core.outer = p.inf_water()
main_cell = openmc.Cell(fill=core)
core_uni = openmc.Universe(cells=[main_cell])
right = openmc.XPlane(x0 = +c.latticePitch19/2,boundary_type=‘vacuum’)
left = openmc.XPlane(x0 = 0,boundary_type=‘reflective’)
down = openmc.YPlane(y0 = -c.latticePitch19/2,boundary_type=‘vacuum’)
up = openmc.YPlane(y0 = 0,boundary_type=‘reflective’)
box=openmc.Cell(region = -right & +left & -up & +down)
box.fill = core_uni
geometry = openmc.Geometry([box])
settings = openmc.Settings()
settings.batches = 100
settings.inactive = 20
settings.particles = 10000
Create an initial uniform spatial source distribution over fissionable zones
lower_left = (-c.latticePitch/2, -c.latticePitch18/2, -1)
upper_right = (c.latticePitch18/2, 0, 1)
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
settings.source = openmc.source.Source(space=uniform_dist)
settings.resonance_scattering = {‘enable’: True, ‘method’: ‘rvs’}
settings.temperature = {‘tolerance’:1000,‘method’:‘interpolation’,‘multipole’:False}
power_mesh = openmc.RegularMesh.from_rect_lattice(core)
power_mesh.dimension = [20,20]
mesh_filter = openmc.MeshFilter(power_mesh)
power = openmc.Tally(name=‘fission power’)
power.filters = [mesh_filter]
power.scores=[‘kappa-fission’]
tallies = openmc.Tallies([power])
model = openmc.model.Model(geometry=geometry,settings=settings,tallies=tallies)
model.export_to_xml()
