Aslak Stubsgaard
GitHub
commit
ac9365798d
17 changed files with 4417323 additions and 3 deletions
29
README.md
29
README.md
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@ -1,7 +1,7 @@
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# msre
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[](https://www.gnu.org/licenses/gpl-3.0)
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> detailed cad model of the [msre](https://en.wikipedia.org/wiki/Molten-Salt_Reactor_Experiment) (molten salt reactor experiment), operated by oak ridge national laboratory 1965-69.
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detailed cad model of the [msre](https://en.wikipedia.org/wiki/Molten-Salt_Reactor_Experiment) (molten salt reactor experiment), operated by oak ridge national laboratory 1965-69.
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## msre core
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@ -11,12 +11,35 @@ the work-in-progress cad model can be found [here](https://cad.onshape.com/docum
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note that this work and the cad model is under the GNU General Public License v3.0
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## prerequisites
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### CAD_to_openMC
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[CAD_to_openMC](https://github.com/openmsr/CAD_to_openMC) is an open-source package to convert CAD geometry (in the form of '.step' files) into an openmc-readable h5m file
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### openmc
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these simulations use [openmc](https://docs.openmc.org/en/stable/). automated source installation scripts for linux can be found [here](https://github.com/openmsr/openmc_install_scripts)
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## simulation guide
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first, clone the repository
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```
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git clone https://github.com/openmsr/are.git
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```
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enter the are folder and run the `run.sh` script
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```
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cd are
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bash run.sh
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```
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## msre heat exchangers
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open-access [master's thesis](https://ltu.diva-portal.org/smash/get/diva2:1546993/FULLTEXT01.pdf) produced by Malcolm Akner about simulations of the heat exchangers of the msre, titled:
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> Validating results from the Molten Salt Reactor Experiment by use of turbulent CFD simulations
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> - A study of a modified U-tube shell-and-tube primary heat exchanger and radiator with molten salts
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Validating results from the Molten Salt Reactor Experiment by use of turbulent CFD simulations
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- A study of a modified U-tube shell-and-tube primary heat exchanger and radiator with molten salts
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### msre primary heat exchanger
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31
run.sh
Executable file
31
run.sh
Executable file
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#!/bin/bash
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set -ex
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PS3='ARE simulations: '
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options=("k eigenvalue" "geometry plot" "neutron flux" "photon flux" "quit")
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select opt in "${options[@]}"
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do
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case $opt in
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"k eigenvalue")
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echo "running k eigenvalue simulation..." &&
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bash ./scripts/k.sh
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;;
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"geometry plot")
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echo "plotting geometry..." &&
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bash ./scripts/gp.sh
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;;
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"neutron flux")
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echo "generating neutron flux plot..." &&
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bash ./scripts/nf.sh
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;;
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"photon flux")
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echo "generating photon flux plot..." &&
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bash ./scripts/pf.sh
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;;
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"quit")
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break
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;;
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*) echo "invalid option $REPLY";;
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esac
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done
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BIN
scripts/__pycache__/materials.cpython-38.pyc
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BIN
scripts/__pycache__/materials.cpython-38.pyc
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Binary file not shown.
10
scripts/gp.sh
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10
scripts/gp.sh
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#!/bin/bash
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set -ex
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if test -f "./h5m_files/ARE.h5m"; then
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python ./scripts/ARE_geometry_plot.py
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else
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python ./scripts/step_to_h5m.py
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python ./scripts/ARE_geometry_plot.py
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fi
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11
scripts/gs.sh
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11
scripts/gs.sh
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#!/bin/bash
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set -ex
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if test -f "./h5m_files/ARE_gamma.h5m"; then
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python ./scripts/ARE_gamma_spectra.py
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else
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python ./step_to_h5m/step_to_h5m_photon.py
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python ./scripts/ARE_gamma_spectra.py
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python ./scripts/ARE_gamma_spectra_plotter.py
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fi
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11
scripts/k.sh
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11
scripts/k.sh
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#!/bin/bash
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set -ex
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if test -f "./h5m_files/msre.h5m"
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then
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python ./scripts/msre.py
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else
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python ./scripts/step_to_h5m.py &&
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python ./scripts/msre.py
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fi
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222
scripts/materials.py
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222
scripts/materials.py
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import sys
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sys.path.append('/opt/openmc/')
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import openmc
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#import neutronsmaterialmaker as nmm
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def define_materials():
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#salt = openmc.Material(name="salt", temperature = 903)
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# salt.add_nuclide('Li7', 10.9566, 'wo')
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# salt.add_element('Be', 6.3492, 'wo')
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# salt.add_element('Zr', 11.1013, 'wo')
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# salt.add_element('Hf', 0.0001, 'wo')
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# salt.add_nuclide('U234', 0.0144, 'wo')
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# salt.add_nuclide('U235', 1.4093, 'wo')
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# salt.add_nuclide('U236', 0.0059, 'wo')
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# salt.add_nuclide('U238', 3.0652, 'wo')
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# salt.add_element('Fe', 0.0162, 'wo')
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# salt.add_element('Cr', 0.0028, 'wo')
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# salt.add_element('Ni', 0.0030, 'wo')
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# salt.add_element('O', 0.0490, 'wo')
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# salt.add_element('F', 67.0270, 'wo')
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# salt.set_density('g/cm3', 2.3223)
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#
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# #fuel salt LiF-BeF2-ZrF4-UF4 (65-29-5-1)
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# salt = openmc.Material(name='salt', temperature = 903)
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# salt.set_density('g/cm3',2.3275)
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# salt.add_nuclide('Li7',65/2)
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# salt.add_element('F',65/2+29*2/3+5*4/5+1*4/5)
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# salt.add_element('Be',29*1/3)
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# salt.add_element('Zr',5*1/5)
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# #salt.add_nuclide('U235',1*1/5*0.93)
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# #salt.add_nuclide('U238',1*1/5*0.07)
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# salt.add_element('U',1*1/5,enrichment=93)
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#Salt definition based on report
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Li = openmc.Material(name='Li', temperature = 903)
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Li.add_nuclide('Li7',1)
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Li.set_density('g/cm3',2.3275)
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Be = openmc.Material(name='Be', temperature = 903)
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Be.add_element('Be',1)
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Be.set_density('g/cm3',2.3275)
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Zr = openmc.Material(name='Zr', temperature = 903)
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Zr.add_element('Zr',1)
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Zr.set_density('g/cm3',2.3275)
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Hf = openmc.Material(name='Hf', temperature = 903)
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Hf.add_element('Hf',1)
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Hf.set_density('g/cm3',2.3275)
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U234 = openmc.Material(name='U234', temperature = 903)
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U234.add_nuclide('U234',1)
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U234.set_density('g/cm3',2.3275)
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U235 = openmc.Material(name='U235', temperature = 903)
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U235.add_nuclide('U235',1)
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U235.set_density('g/cm3',2.3275)
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U236 = openmc.Material(name='U236', temperature = 903)
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U236.add_nuclide('U236',1)
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U236.set_density('g/cm3',2.3275)
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U238 = openmc.Material(name='U238', temperature = 903)
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U238.add_nuclide('U238',1)
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U238.set_density('g/cm3',2.3275)
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Fe = openmc.Material(name='Fe', temperature = 903)
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Fe.add_element('Fe',1)
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Fe.set_density('g/cm3',2.3275)
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Cr = openmc.Material(name='Cr', temperature = 903)
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Cr.add_element('Cr',1)
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Cr.set_density('g/cm3',2.3275)
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Ni = openmc.Material(name='Ni', temperature = 903)
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Ni.add_element('Ni',1)
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Ni.set_density('g/cm3',2.3275)
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O = openmc.Material(name='O', temperature = 903)
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O.add_element('O',1)
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O.set_density('g/cm3',2.3275)
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F = openmc.Material(name='F', temperature = 903)
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F.add_element('F',1)
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F.set_density('g/cm3',2.3275)
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salt = openmc.Material.mix_materials(
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[Li, Be, Zr, Hf, U234, U235, U236, U238, Fe, Cr, Ni, O, F],
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[10.9566/100, 6.3492/100, 11.1013/100, 0.0001/100, 0.0144/100, 1.4093/100, 0.0059/100, 3.0652/100, 0.0162/100, 0.0028/100, 0.0030/100, 0.0490/100, 67.0270/100],
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'wo')
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salt.name="salt"
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#moderator blocks170
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graphite = openmc.Material(name='graphite',temperature=903)
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graphite.set_density('g/cm3',1.86)
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graphite.add_element('C',1)
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graphite.add_s_alpha_beta('c_Graphite')
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#inor
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Mo = openmc.Material(name='Mo', temperature = 903)
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Mo.add_element('Mo',1)
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Mo.set_density('g/cm3',2.3275)
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C = openmc.Material(name='C', temperature = 903)
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C.add_element('Mo',1)
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C.set_density('g/cm3',2.3275)
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Al = openmc.Material(name='Al', temperature = 903)
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Al.add_element('Al',1)
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Al.set_density('g/cm3',2.3275)
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Ti = openmc.Material(name='Ti', temperature = 903)
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Ti.add_element('Ti',1)
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Ti.set_density('g/cm3',2.3275)
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S = openmc.Material(name='S', temperature = 903)
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S.add_element('S',1)
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S.set_density('g/cm3',2.3275)
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Mn = openmc.Material(name='Mn', temperature = 903)
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Mn.add_element('Mn',1)
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Mn.set_density('g/cm3',2.3275)
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Si = openmc.Material(name='Si', temperature = 903)
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Si.add_element('Si',1)
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Si.set_density('g/cm3',2.3275)
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Cu = openmc.Material(name='Cu', temperature = 903)
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Cu.add_element('Cu',1)
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Cu.set_density('g/cm3',2.3275)
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B = openmc.Material(name='B', temperature = 903)
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B.add_element('B',1)
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B.set_density('g/cm3',2.3275)
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W = openmc.Material(name='W', temperature = 903)
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W.add_element('W',1)
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W.set_density('g/cm3',2.3275)
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P = openmc.Material(name='P', temperature = 903)
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P.add_element('P',1)
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P.set_density('g/cm3',2.3275)
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Co = openmc.Material(name='Co', temperature = 903)
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Co.add_element('Co',1)
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Co.set_density('g/cm3',2.3275)
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inor = openmc.Material.mix_materials(
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[Ni, Mo, Cr, Fe, C, Al, Ti, S, Mn, Si, Cu, B, W, P, Co],
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[68/100, 17/100, 7/100, 5/100, 0.06/100, 0.2/100, 0.2/100, 0.02/100, 0.8/100, 0.8/100, 0.3/100, 0.01/100, 0.4/100, 0.015/100, 0.195/100],
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'wo')
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inor.name="inor"
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inor.set_density('g/cm3',8.7745)
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# inor = openmc.Material(name='inor',temperature=903)
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# inor.set_density('g/cm3',8.7745)
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# inor.add_element('Ni',(66+71)/2,'wo')
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# inor.add_element('Mo',(15+18)/2,'wo')
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# inor.add_element('Cr',(6+8)/2,'wo')
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# inor.add_element('Fe',5,'wo')
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# inor.add_element('C',(0.04+0.08)/2,'wo')
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# inor.add_element('Al',0.25,'wo')
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# inor.add_element('Ti',0.25,'wo')
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# inor.add_element('S',0.02,'wo')
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# inor.add_element('Mn',1.0,'wo')
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# inor.add_element('Si',1.0,'wo')
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# inor.add_element('Cu',0.35,'wo')
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# inor.add_element('B',0.010,'wo')
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# inor.add_element('W',0.5,'wo')
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# inor.add_element('P',0.015,'wo')
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# inor.add_element('Co',0.2,'wo')
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#helium
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helium = openmc.Material(name='helium')
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helium.add_element('He',1.0)
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helium.set_density('g/cm3',1.03*(10**-4))
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#Control rods inconel clad
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trace = 0.01
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inconel = openmc.Material(name='inconel', temperature = 903)
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inconel.add_element('Ni',78.5,percent_type='wo')
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inconel.add_element('Cr',14.0,percent_type='wo')
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inconel.add_element('Fe',6.5,percent_type='wo')
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inconel.add_element('Mn',0.25,percent_type='wo')
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inconel.add_element('Si',0.25,percent_type='wo')
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inconel.add_element('Cu',0.2,percent_type='wo')
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inconel.add_element('Co',0.2,percent_type='wo')
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inconel.add_element('Al',0.2,percent_type='wo')
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inconel.add_element('Ti',0.2,percent_type='wo')
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inconel.add_element('Ta',0.5,percent_type='wo')
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inconel.add_element('W',0.5,percent_type='wo')
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inconel.add_element('Zn',0.2,percent_type='wo')
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inconel.add_element('Zr',0.1,percent_type='wo')
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inconel.add_element('C',trace,percent_type='wo')
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inconel.add_element('Mo',trace,percent_type='wo')
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inconel.add_element('Ag',trace,percent_type='wo')
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inconel.add_element('B',trace,percent_type='wo')
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inconel.add_element('Ba',trace,percent_type='wo')
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inconel.add_element('Be',trace,percent_type='wo')
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inconel.add_element('Ca',trace,percent_type='wo')
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inconel.add_element('Cd',trace,percent_type='wo')
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inconel.add_element('V',trace,percent_type='wo')
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inconel.add_element('Sn',trace,percent_type='wo')
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inconel.add_element('Mg',trace,percent_type='wo')
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inconel.set_density('g/cm3',8.5)
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#Control rods bushing posion material
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Gd2O3 = openmc.Material()
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Gd2O3.add_element('Gd',2)
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Gd2O3.add_element('O',3)
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Gd2O3.set_density('g/cm3',5.873)
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Al2O3 = openmc.Material()
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Al2O3.add_element('Al',2)
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Al2O3.add_element('O',3)
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Al2O3.set_density('g/cm3',5.873)
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bush = openmc.Material.mix_materials([Gd2O3,Al2O3],[0.7,0.3],'wo')
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bush.name='bush'
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#Concrete block
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#concrete=nmm.Material.from_library('Concrete, Regular')
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#concrete=concrete.openmc_material
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#concrete.name="concrete"
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#Baryte block
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baryte = openmc.Material(name="baryte", temperature = 903)
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baryte.add_element('Ba',1)
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baryte.add_element('S',1)
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baryte.add_element('O',4)
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baryte.set_density('g/cm3',4.5)
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#Thermal shielding
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water = openmc.Material()
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water.add_element('H',2)
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water.add_element('O',1)
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water.set_density('g/cm3',0.997)
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#steel = nmm.Material.from_library('Steel, Carbon')
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#steel=steel.openmc_material
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#shield = openmc.Material.mix_materials([water,steel],[0.5,0.5],'wo')
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#shield.name='waterSteel'
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detector = openmc.Material(name = 'water')
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detector.add_element('O',76.2,percent_type='wo')
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detector.add_element('C',11.1,percent_type='wo')
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detector.add_element('H',10.1,percent_type='wo')
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detector.add_element('N',2.6,percent_type='wo')
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detector.set_density('g/cm3',1.0)
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mats = openmc.Materials([salt,graphite,inor,helium,inconel,bush])#,concrete,baryte,shield,detector])
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mats.export_to_xml()
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return mats
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36
scripts/msre.py
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36
scripts/msre.py
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import matplotlib.pyplot as plt
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from materials import *
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import openmc
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###############################################################################
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#k eigenvalue simulation on ARE (all safety rods fully inserted)
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###############################################################################
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h5m_filepath = 'h5m_files/msre.h5m'
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#Materials
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define_materials()
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#geometry
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graveyard=openmc.Sphere(r=10000,boundary_type='vacuum')
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cad_univ = openmc.DAGMCUniverse(filename=h5m_filepath,auto_geom_ids=True,universe_id=996 )
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cad_cell = openmc.Cell(cell_id=997 , region= -graveyard, fill= cad_univ)
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root = openmc.Universe(universe_id=998)
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root.add_cells([cad_cell])
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geometry = openmc.Geometry(root)
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geometry.export_to_xml()
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settings = openmc.Settings()
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settings.temperature = {'method':'interpolation'}
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settings.batches = 100
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settings.inactive = 10
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settings.particles = 10000
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settings.max_lost_particles = 10000
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source_area = openmc.stats.Box([-200., -200., -200.],[ 200., 200., 200.],only_fissionable = True)
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settings.source = openmc.Source(space=source_area)
|
||||
settings.export_to_xml()
|
||||
|
||||
openmc.run()
|
||||
57
scripts/msre_geometry_plot.py
Normal file
57
scripts/msre_geometry_plot.py
Normal file
|
|
@ -0,0 +1,57 @@
|
|||
import matplotlib
|
||||
import openmc
|
||||
from materials import *
|
||||
|
||||
###############################################################################
|
||||
#generate geometry plot of are (all safety rods fully inserted)
|
||||
###############################################################################
|
||||
|
||||
#geometry
|
||||
h5m_filepath = '../h5m_files/msre_simp_cubit/msre_simp_1e-2.h5m'
|
||||
graveyard = openmc.Sphere(r=10000,boundary_type='vacuum')
|
||||
cad_univ = openmc.DAGMCUniverse(filename=h5m_filepath,auto_geom_ids=True)
|
||||
cad_cell = openmc.Cell(region=-graveyard,fill=cad_univ)
|
||||
root=openmc.Universe()
|
||||
root.add_cells([cad_cell])
|
||||
geometry=openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
# materials
|
||||
mats = define_materials()
|
||||
#mats = openmc.Materials([salt,BeO,inconel,insulation,coolant,helium,stainless,boron])
|
||||
mats.export_to_xml()
|
||||
|
||||
#plotting geometry
|
||||
plots = openmc.Plots()
|
||||
|
||||
x_width = 300
|
||||
y_width = 300
|
||||
|
||||
#xy plot
|
||||
p1 = openmc.Plot()
|
||||
p1.width = (x_width,y_width)
|
||||
p1.origin = (0,0,100)
|
||||
p1.pixels = (1000, 1000)
|
||||
p1.color_by = 'material'
|
||||
|
||||
#xz plot (split plane)
|
||||
p2 = openmc.Plot()
|
||||
p2.basis = 'xz'
|
||||
p2.origin = (0,0,130)
|
||||
p2.width = (x_width,y_width)
|
||||
p2.pixels = (1000, 1000)
|
||||
p2.color_by = 'material'
|
||||
|
||||
p3 = openmc.Plot()
|
||||
p3.basis = 'yz'
|
||||
p3.origin = (0,0,130)
|
||||
p3.width = (x_width,y_width)
|
||||
p3.pixels = (1000, 1000)
|
||||
p3.color_by = 'material'
|
||||
|
||||
plots.append(p1)
|
||||
plots.append(p2)
|
||||
plots.append(p3)
|
||||
plots.export_to_xml()
|
||||
|
||||
openmc.plot_geometry()
|
||||
68
scripts/msre_neutron_flux.py
Normal file
68
scripts/msre_neutron_flux.py
Normal file
|
|
@ -0,0 +1,68 @@
|
|||
import matplotlib.pyplot as plt
|
||||
import openmc
|
||||
from materials import *
|
||||
|
||||
###############################################################################
|
||||
#create .png file of neutron flux (all safety rods inserted)
|
||||
###############################################################################
|
||||
|
||||
#Geometry
|
||||
h5m_filepath = '../h5m_files/msre_simp_cubit/msre_simp_1e-2.h5m'
|
||||
graveyard = openmc.Sphere(r=10000,boundary_type='vacuum')
|
||||
cad_univ = openmc.DAGMCUniverse(filename=h5m_filepath,auto_geom_ids=True)
|
||||
cad_cell = openmc.Cell(region=-graveyard,fill=cad_univ)
|
||||
root=openmc.Universe()
|
||||
root.add_cells([cad_cell])
|
||||
geometry=openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
#materials
|
||||
mats=define_materials()
|
||||
#mats = openmc.Materials([salt,BeO,inconel,insulation,coolant,helium,stainless,boron,blanket,shield])
|
||||
mats.export_to_xml()
|
||||
|
||||
settings = openmc.Settings()
|
||||
settings.temperature = {'method':'interpolation'}
|
||||
settings.batches = 100
|
||||
settings.inactive = 10
|
||||
settings.particles = 5000
|
||||
settings.max_lost_particles = 10000
|
||||
source_area = openmc.stats.Box([-200., -200., -200.],[ 200., 200., 200.],only_fissionable = True)
|
||||
settings.source = openmc.Source(space=source_area)
|
||||
settings.export_to_xml()
|
||||
|
||||
#tallies
|
||||
tallies = openmc.Tallies()
|
||||
|
||||
mesh = openmc.RegularMesh()
|
||||
mesh.dimension = [1000,1000]
|
||||
mesh.lower_left = [-300,-300]
|
||||
mesh.upper_right = [300,300]
|
||||
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
tally = openmc.Tally(name='flux')
|
||||
tally.filters = [mesh_filter]
|
||||
tally.scores = ['flux','fission']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
model = openmc.model.Model(geometry, mats, settings, tallies)
|
||||
sp_filename = model.run()
|
||||
sp = openmc.StatePoint(sp_filename)
|
||||
s_tally = sp.get_tally(scores=['flux','fission'])
|
||||
|
||||
flux = s_tally.get_slice(scores=['flux'])
|
||||
fission = s_tally.get_slice(scores=['fission'])
|
||||
|
||||
flux.std_dev.shape = (1000,1000)
|
||||
flux.mean.shape = (1000,1000)
|
||||
fission.std_dev.shape = (1000,1000)
|
||||
fission.mean.shape = (1000,1000)
|
||||
|
||||
fig = plt.subplot(121)
|
||||
fig.axis([350,650,350,650])
|
||||
fig.pixels = (2000,2000)
|
||||
fig.imshow(flux.mean)
|
||||
plt.savefig('neutron_flux', dpi=2000)
|
||||
74
scripts/msre_photon_flux.py
Normal file
74
scripts/msre_photon_flux.py
Normal file
|
|
@ -0,0 +1,74 @@
|
|||
import matplotlib.pyplot as plt
|
||||
import openmc
|
||||
from materials import *
|
||||
|
||||
###############################################################################
|
||||
#create .png file of photon flux (all safety rods inserted)
|
||||
###############################################################################
|
||||
|
||||
#Geometry
|
||||
h5m_filepath = 'h5m_files/ARE.h5m'
|
||||
graveyard = openmc.Sphere(r=10000,boundary_type='vacuum')
|
||||
cad_univ = openmc.DAGMCUniverse(filename=h5m_filepath,auto_geom_ids=True)
|
||||
cad_cell = openmc.Cell(region=-graveyard,fill=cad_univ)
|
||||
root=openmc.Universe()
|
||||
root.add_cells([cad_cell])
|
||||
geometry=openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
#materials
|
||||
mats = openmc.Materials([salt,BeO,inconel,insulation,coolant,helium,stainless,boron,blanket,shield])
|
||||
mats.export_to_xml()
|
||||
|
||||
settings = openmc.Settings()
|
||||
settings.temperature = {'method':'interpolation'}
|
||||
settings.batches = 100
|
||||
settings.inactive = 10
|
||||
settings.particles = 5000
|
||||
settings.photon_transport = True
|
||||
source_area = openmc.stats.Box([-200., -200., -200.],[ 200., 200., 200.],only_fissionable = True)
|
||||
settings.source = openmc.Source(space=source_area)
|
||||
settings.export_to_xml()
|
||||
|
||||
#tallies
|
||||
tallies = openmc.Tallies()
|
||||
|
||||
# sets up filters for the tallies
|
||||
photon_particle_filter = openmc.ParticleFilter(['photon']) # note the use of photons here
|
||||
energy_bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
|
||||
energy_filter = openmc.EnergyFilter(energy_bins)
|
||||
|
||||
|
||||
mesh = openmc.RegularMesh()
|
||||
mesh.dimension = [1000,1000]
|
||||
mesh.lower_left = [-300,-300]
|
||||
mesh.upper_right = [300,300]
|
||||
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
tally = openmc.Tally(name='flux')
|
||||
tally.filters = [mesh_filter,photon_particle_filter]
|
||||
tally.scores = ['flux']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
# combine all the required parts to make a model
|
||||
model = openmc.model.Model(geom, mats, settings, tallies)
|
||||
|
||||
# remove old files and runs OpenMC
|
||||
results_filename = model.run()
|
||||
|
||||
sp = openmc.StatePoint('statepoint.100.h5')
|
||||
s_tally = sp.get_tally(scores=['flux'])
|
||||
|
||||
flux = s_tally.get_slice(scores=['flux'])
|
||||
|
||||
flux.std_dev.shape = (1000,1000)
|
||||
flux.mean.shape = (1000,1000)
|
||||
|
||||
fig = plt.subplot(121)
|
||||
fig.axis([350,650,350,650])
|
||||
fig.pixels = (2000,2000)
|
||||
fig.imshow(flux.mean)
|
||||
plt.savefig('photon_flux', dpi=2000)
|
||||
10
scripts/nf.sh
Normal file
10
scripts/nf.sh
Normal file
|
|
@ -0,0 +1,10 @@
|
|||
#!/bin/bash
|
||||
|
||||
set -ex
|
||||
|
||||
if test -f "./h5m_files/ARE.h5m"; then
|
||||
python ./scripts/ARE_neutron_flux.py
|
||||
else
|
||||
python ./scripts/step_to_h5m.py
|
||||
python ./scripts/ARE_neutron_flux.py
|
||||
fi
|
||||
10
scripts/pf.sh
Normal file
10
scripts/pf.sh
Normal file
|
|
@ -0,0 +1,10 @@
|
|||
#!/bin/bash
|
||||
|
||||
set -ex
|
||||
|
||||
if test -f "./h5m_files/ARE.h5m"; then
|
||||
python ./scripts/ARE_photon_flux.py
|
||||
else
|
||||
python ./scripts/step_to_h5m.py
|
||||
python ./scripts/ARE_photon_flux.py
|
||||
fi
|
||||
29
scripts/step_to_h5m.py
Normal file
29
scripts/step_to_h5m.py
Normal file
|
|
@ -0,0 +1,29 @@
|
|||
###############################################################################
|
||||
# Converting step files to h5m file to be read by openmc
|
||||
###############################################################################
|
||||
import numpy as np
|
||||
import os
|
||||
import sys
|
||||
sys.path.append('../CAD_to_openMC/src/')
|
||||
import CAD_to_OpenMC.assembly as ab
|
||||
###############################################################################
|
||||
|
||||
# inputs
|
||||
step_filepath = "./step_files/msre.step"
|
||||
h5m_out_filepath = os.getcwd() + '/h5m_files/msre.h5m'
|
||||
|
||||
# mesher config
|
||||
ab.mesher_config['min_mesh_size'] = 0.2
|
||||
ab.mesher_config['mesh_algorithm'] = 4
|
||||
ab.mesher_config['threads'] = 6
|
||||
#ab.mesher_config['max_mesh_size'] = 1000
|
||||
#ab.mesher_config['curve_samples'] = 100
|
||||
#ab.mesher_config['vetoed'] = [474,1157,1243,1341,1537]
|
||||
#ab.mesher_config['angular_tolerance'] = 0.09
|
||||
#ab.mesher_config['tolerance'] = 0.001
|
||||
|
||||
# output
|
||||
a=ab.Assembly()
|
||||
a.stp_files=[step_filepath]
|
||||
a.import_stp_files()
|
||||
a.solids_to_h5m(backend='gmsh',h5m_filename=h5m_out_filepath)
|
||||
1345087
step_files/msre.step
Normal file
1345087
step_files/msre.step
Normal file
File diff suppressed because it is too large
Load diff
1539188
step_files/msre_no_cr.step
Normal file
1539188
step_files/msre_no_cr.step
Normal file
File diff suppressed because it is too large
Load diff
1532453
step_files/msre_simp.step
Normal file
1532453
step_files/msre_simp.step
Normal file
File diff suppressed because it is too large
Load diff
Loading…
Add table
Add a link
Reference in a new issue