Need help to model the same problem in OpenMC

User Support
1

Hi, I am new to use OpenMC, and want to learn by example, I have the problem modeled in Serpent and want to model the same with OpenMC, please write the OpenMC version of the given input file

Serpent version:
set title “cylMCFR radius 19.0, height 100.0, reflector 25.0”
%surface definitions____________________
surf 1 cylz 0.0 0.0 19.0 % fuel
surf 2 cylz 0.0 0.0 44.0 % reflector

%cell definitions_______________________
cell 30 re refl 1 -2 % radial reflector
cell 80 0 fill re 1 -2
cell 50 fs fuelsalt -1 % fuel salt
cell 85 0 fill fs -1
cell 99 0 outside 2

% Fuel salt: 66.66%NaCl + 33.34%UCl3, U enrichment 0.93
mat fuelsalt -3.16391796 rgb 240 30 30 burn 1 tmp 900.000
11023.09c -0.100274788825 % Na-23
17035.09c -0.289847995589 % Cl-35
17037.09c -0.096758222661 % Cl-37
92234.09c -0.004225929437 % U-234
92235.09c -0.476858378996 % U-235
92236.09c -0.002202896269 % U-236
92238.09c -0.029831788223 % U-238

% MgO reflector
mat refl -3.5 tmp 873.0 rgb 75 75 75
12024.06c 1.0
8016.06c 1.0

set mvol fuelsalt 0 226822.98958918

set pop 100000 240 40 % N pop and criticality cycles
set power 300000.0 % Power, 300 thermal kW
set bc 1 % vacuum

ene six_group 1 1E-11 7.485E-4 5.5308E-3 2.47875E-2 0.4979 2.2313 12
set gcu re fs % group constant generation in unis fs and re
set nfg six_group
set micro nj20

det default n
dn 1
0 44 25 % 25 radial bins, 0 to 44
0 360 1 % 1 angual bin, 0 to 360 deg
-25 125 75 % 75 axial bin, -25 to 125
de six_group

% Data Libraries
set acelib “/opt/JEFF-3.3/sss_jeff33.xsdir”
set declib “/opt/JEFF-3.3/jeff33.dec”
set nfylib “/opt/JEFF-3.3/jeff33.nfy”

% Plots
plot 3 1500 1500
plot 2 1500 1500

2

One simple option is to use serpent openmc convertor.

3

Hi Habib, welcome to the openmc community.
You could try the serpent adapter mentioned by Khurrum since I think the serpent geometry you want to model is still a simple model with cylz surfaces. But if the geometry becomes complicated, I recommend you build your geometry manually from the surface to the cell and universe based on your serpent model. That way you can avoid geometry errors.

4

The converter currently only handles geometry and material information, while i have model the same problem for simple criticality calculations but the problem is that i got very high relative errer of (817.22 pcm) keff_serpent = 1.09934, keff_openmc = 1.090356, i want to please check my input file and correctme if there are some mistakes

OpenMC Input File:

import numpy as np
import matplotlib.pyplot as plt
from PIL import Image # For image conversion
import openmc

Step 1: Define geometry dimensions

fuel_radius = 19.0 # Fuel radius in cm
reflector_thickness = 25.0 # Reflector thickness in cm
total_radius = 44.0
height = 100.0

Step 2: Define materials

Fuel Salt Material (NaCl-UCl3 mixture)

fuel = openmc.Material(name=‘Fuel Salt’)
fuel.add_element(‘Na’, 0.100274788825, percent_type=‘wo’)
fuel.add_nuclide(‘Cl35’, 0.289847995589, percent_type=‘wo’)
fuel.add_nuclide(‘Cl37’, 0.096758222661, percent_type=‘wo’)
fuel.add_nuclide(‘U234’, 0.004225929437, percent_type=‘wo’)
fuel.add_nuclide(‘U235’, 0.476858378996, percent_type=‘wo’)
fuel.add_nuclide(‘U236’, 0.002202896269, percent_type=‘wo’)
fuel.add_nuclide(‘U238’, 0.029831788223, percent_type=‘wo’)
fuel.set_density(‘g/cm3’, 3.16391796)
fuel.temperature = 900 # Set current temperature

Reflector Material (Magnesium Oxide)

reflector = openmc.Material(name=‘Magnesium Oxide’)
reflector.add_element(‘Mg’, 1.0) # Magnesium
reflector.add_element(‘O’, 1.0) # Oxygen
reflector.set_density(‘g/cm3’, 3.5)
reflector.temperature = 873.0 # Set current temperature

Step 3: Create a materials collection and export to XML

materials = openmc.Materials([fuel, reflector])
materials.export_to_xml()

Step 3: Define geometry

Fuel cylinder (infinite along z-axis)

fuel_cylinder = openmc.ZCylinder(r=fuel_radius)

Reflector cylinder (infinite along z-axis)

reflector_cylinder = openmc.ZCylinder(r=total_radius, boundary_type=‘vacuum’)

Define planes for z-boundaries

bottom_plane = openmc.ZPlane(z0=-0.5 * height, boundary_type=‘reflective’)
top_plane = openmc.ZPlane(z0=0.5 * height, boundary_type=‘reflective’)

Define regions

fuel_region = -fuel_cylinder & +bottom_plane & -top_plane # Inside the fuel cylinder
reflector_region = +fuel_cylinder & -reflector_cylinder & +bottom_plane & -top_plane # Between fuel and reflector cylinders

Assign materials to regions

fuel_cell = openmc.Cell(region=fuel_region, fill=fuel)
reflector_cell = openmc.Cell(region=reflector_region, fill=reflector)

Create a universe

root_universe = openmc.Universe(cells=[fuel_cell, reflector_cell])

Step 4: Export geometry

geometry = openmc.Geometry(root_universe)
geometry.export_to_xml()

Step 5: Define settings

settings = openmc.Settings()
settings.particles = 100000
settings.batches = 240
settings.inactive = 40
settings.output = {‘tallies’: True}

Define the bounds for the uniform source distribution

bounds = [-44, -44, -50, 44, 44, 50]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings.source = openmc.Source(space=uniform_dist)
settings.trigger_active = True
settings.trigger_max_batches = settings.batches * 4
settings.temperature = {‘method’: ‘interpolation’, ‘multipole’: True}
settings.export_to_xml()

Step 6: Run simulation

openmc.run()

5

Have you tried to run using the same JEFF-3.3 nuclear data library?
I ran your input with ENDF VII.1 and I got k-eff 1.090xx, and sorry I didn’t try to run it with the JEFF 3.3 lib even though we can download it from Other Data Libraries | OpenMC.
I think your openmc model is consistent even though you use 100 cm height reflective BC, but I have tried to run without it and it was still around 1.090. I also checked my endf lib, and I didn’t find any s(alpha,beta) for MgO.

I hope other members can give their recommendations on this code-to-code case.