Candu Reactor compared k_eff values with Serpent

[emilmammadzada]

TL;DR

Candu Reactor compared k_eff values with Serpent

Dear experts, I am trying to set the Serpent input file for the candu reactor in Openmc accordingly. But in the output section, Serpent k_eff values are around 0.9, while Openmc shows 0.42. How can I solve this problem or what are the errors in the files I added below that cause this difference?
[from math import pi, sin, cos
import numpy as np
import openmc
import json
import numpy as np
from pathlib import Path
import openmc.deplete
# Fuel and clad
temperature_fuel = 960 #Fuel at 960 K with a density of 10.52 g/cc
temperature_structure = 345.15 #Moderator at 345.15 K with a density of 1.086 g/cc
r_fuel = 0.6122
r_clad = 0.6540
pressure_tube_ir = 5.16890
pressure_tube_or = 5.60320
calendria_ir = 6.44780
calendria_or = 6.58750
# Fuel
fuel = openmc.Material(name='fuel', temperature=temperature_fuel)
fuel.add_nuclide('U235', 6.27118E-1 , 'wo')
fuel.add_nuclide('U238', 8.75256E+1 , 'wo')
fuel.add_nuclide('O16', 1.18473E+1 , 'wo')
fuel.volume = 37*pi*r_fuel**2
fuel.set_density('g/cm3', 10.4375010)
# Coolant Water
heavy_water_coolant = openmc.Material(name='heavy water coolant')
heavy_water_coolant.add_nuclide('H2', 1.99768E-1 , 'wo' )
heavy_water_coolant.add_nuclide('O16', 7.99449E-1 , 'wo' )
heavy_water_coolant.add_nuclide('H1', 7.83774E-4 , 'wo' )
heavy_water_coolant.set_density('g/cm3', 0.812120)
# Calandria tube
clad_calandria = openmc.Material(name='Calandria')
clad_calandria.add_nuclide('Mn55', 1.60000E-1 , 'wo' )
clad_calandria.add_element('Ni', 6.00000E-2 , 'wo' )
clad_calandria.add_element('Cr', 1.10000E-1 , 'wo' )
clad_calandria.add_element('Zr', 9.97100E+1 , 'wo' )
clad_calandria.add_nuclide('B10', 5.7409e-05 , 'wo' )
clad_calandria.add_nuclide('B11', 2.5259E-04 , 'wo' )
clad_calandria.set_density('g/cm3', 6.44)
# Clad
clad = openmc.Material(name='Clad')
clad.add_nuclide('Mn55', 1.60000E-1 , 'wo' )
clad.add_element('Ni', 6.00000E-2 , 'wo' )
clad.add_element('Cr', 1.10000E-1 , 'wo' )
clad.add_element('Zr', 9.97100E+1 , 'wo' )
clad.add_nuclide('B10', 5.7409e-05 , 'wo' )
clad.add_nuclide('B11', 2.5259E-04 , 'wo' )
clad.set_density('g/cm3', 6.44)
# Moder
heavy_water = openmc.Material(name='heavy water')
heavy_water.add_nuclide('H2', 2.01016E-1, 'wo' )
heavy_water.add_nuclide('O16', 7.98895E-1, 'wo' )
heavy_water.add_nuclide('H1', 8.96000E-5, 'wo' )
heavy_water.set_density('g/cm3', 1.082885 )
heavy_water.add_s_alpha_beta('c_D_in_D2O')
# Pressure Tube
tube = openmc.Material(name='PressTube')
tube.add_element('Zr', 9.75000E+1 , 'wo' )
tube.add_nuclide('B10', 3.8889E-05 , 'wo' )
tube.add_nuclide('B11', 1.7111E-04 , 'wo' )
tube.set_density('g/cm3', 6.57)
mats = openmc.Materials([fuel, heavy_water_coolant, clad_calandria, clad, heavy_water, tube ])
mats.export_to_xml()

# Radius to center of each ring of fuel pins
ring_radii = np.array([0.0, 1.4885, 2.8755, 4.3305])
# These are the surfaces that will divide each of the rings
radial_surf = [openmc.ZCylinder(R=r) for r in
(ring_radii[:-1] + ring_radii[1:])/2]

water_cells = []
for i in range(ring_radii.size):
# Create annular region
if i == 0:
water_region = -radial_surf
elif i == ring_radii.size - 1:
water_region = +radial_surf[i-1]
else:
water_region = +radial_surf[i-1] & -radial_surf

water_cells.append(openmc.Cell(fill=heavy_water, region=water_region))
bundle_universe = openmc.Universe(cells=water_cells)
surf_fuel = openmc.ZCylinder(R=r_fuel)
surf_clad = openmc.ZCylinder(R=r_clad)

fuel_cell = openmc.Cell(fill=fuel, region=-surf_fuel)
clad_cell = openmc.Cell(fill=clad, region=+surf_fuel & -surf_clad)
cool_cell = openmc.Cell(fill=heavy_water_coolant, region=+surf_clad & -surf_clad)

pin_universe = openmc.Universe(cells=(fuel_cell, clad_cell, cool_cell))
num_pins = [1, 6, 12, 18]
angles = [0, 0, 15, 0]

for i, (r, n, a) in enumerate(zip(ring_radii, num_pins, angles)):
for j in range(n):
# Determine location of center of pin
theta = (a + j/n*360.) * pi/180.
x = r*cos(theta)
y = r*sin(theta)

pin_boundary = openmc.ZCylinder(x0=x, y0=y, R=r_clad)
water_cells.region &= +pin_boundary


# That we can identify the pin later when looking at tallies
pin = openmc.Cell(fill=pin_universe, region=-pin_boundary)
pin.translation = (x, y, 0)
pin.id = (i + 1)*100 + j
bundle_universe.add_cell(pin)

pt_inner = openmc.ZCylinder(R=pressure_tube_ir)
pt_outer = openmc.ZCylinder(R=pressure_tube_or)
calendria_inner = openmc.ZCylinder(R=calendria_ir)
calendria_outer = openmc.ZCylinder(R=calendria_or,boundary_type='reflective')

bundle = openmc.Cell(fill=bundle_universe, region=-pt_inner)
pressure_tube = openmc.Cell(fill=tube, region=+pt_inner & -pt_outer)
v1 = openmc.Cell(region=+pt_outer & -calendria_inner)
calendria = openmc.Cell(fill=clad_calandria, region=+calendria_inner & -calendria_outer)

root_universe = openmc.Universe(cells=[bundle, pressure_tube, v1, calendria])

geom = openmc.Geometry(root_universe)
geom.export_to_xml()
settings = openmc.Settings()
settings = openmc.Settings()
settings.particles = 500
settings.batches = 50
settings.inactive = 5
settings.source = openmc.Source(space=openmc.stats.Point())
settings.temperature = {
'default': temperature_structure,
'method': 'interpolation',
}
settings.export_to_xml()
openmc.run()]