Merge pull request #41 from fusion-energy/develop

updates to work with latest openmc compile
fusion-energy · May 31, 2022 · ffd4ab4 · ffd4ab4
2 parents 13b9d0c + 4f96c54
commit ffd4ab4
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Showing 6 changed files with 56 additions and 60 deletions.
diff --git a/Dockerfile b/Dockerfile
@@ -32,7 +32,7 @@
 # docker build -t fusion-neutronics-workflow:embree --build-arg compile_cores=7 --build-arg build_double_down=ON --build-arg include_avx=true .
 
 # base image contains nuclear data
-FROM ghcr.io/openmc-data-storage/miniconda3_4.9.2_endfb-7.1_nndc_tendl_2019:latest
+FROM continuumio/miniconda3:4.10.3
 
 ARG compile_cores=1
 ARG include_avx=true
@@ -163,7 +163,7 @@ RUN if [ "$build_double_down" = "ON" ] ; \
 # Clone and install DAGMC
 RUN mkdir DAGMC && \
     cd DAGMC && \
-    git clone --single-branch --branch v3.2.1 --depth 1 https://github.com/svalinn/DAGMC.git && \
+    git clone --single-branch --branch v3.2.2 --depth 1 https://github.com/svalinn/DAGMC.git && \
     mkdir build && \
     cd build && \
     cmake ../DAGMC -DBUILD_TALLY=ON \
@@ -192,26 +192,18 @@ RUN cd /opt && \
     cd ..  && \
     pip install -e .[test]
 
+COPY environment.yml .
+RUN conda env update -f environment.yml
 
-RUN conda install -c fusion-energy -c cadquery -c conda-forge paramak_develop
+# downloads nuclear data from zip file an uncompress
+# RUN download_nndc
 
-# pip installs latest versions of analysis packages
-RUN pip install openmc-dagmc-wrapper
-RUN pip install openmc-plasma-source
-RUN pip install openmc_tally_unit_converter
-RUN pip install spectrum_plotter
-RUN pip install regular_mesh_plotter
-RUN pip install openmc_data_downloader
-
-# solves binary incompatabilitiy error
-RUN pip install numpy --upgrade
-
-
-# # installs python packages and nuclear data (quick as it does not overwrite existing h5 files)
-# RUN openmc_data_downloader -d nuclear_data -e all -i H3 -l ENDFB-7.1-NNDC TENDL-2019 -p neutron photon --no-overwrite
+# this is an optional way to install python packages and nuclear data (quick as it does not overwrite existing h5 files)
+RUN openmc_data_downloader -d nuclear_data -e all -i H3 -l ENDFB-7.1-NNDC TENDL-2019 -p neutron photon --no-overwrite
 
 # setting enviromental varibles
 ENV OPENMC_CROSS_SECTIONS=/nuclear_data/cross_sections.xml
+# ENV OPENMC_CROSS_SECTIONS=/nndc-b7.1-hdf5/endfb71_hdf5/cross_sections.xml
 ENV PATH="/MOAB/build/bin:${PATH}"
 ENV PATH="/DAGMC/bin:${PATH}"
 

diff --git a/README.md b/README.md
@@ -89,14 +89,10 @@ Links to the packages that are utilized by the fusion-neutronics-workflow
 
 * Open source projects created and maintained
 
-    * [Paramak](https://github.com/fusion-energy/openmc_data_downloader) -
+    * [Paramak](https://github.com/fusion-energy/paramak) -
     automated production of fusion reactor CAD models (stp and stl files) from
     parameters.
 
-    * [cad_to_h5m](https://github.com/fusion-energy/cad_to_h5m) automated
-    conversion of STP or SAT CAD files to h5m files compatible with DAGMC
-    enabled particle transport codes.
-
     * [stl_to_h5m](https://github.com/fusion-energy/stl_to_h5m) automated
     conversion of STL files to h5m files compatible with DAGMC enabled
     particle transport codes.
@@ -112,8 +108,9 @@ Links to the packages that are utilized by the fusion-neutronics-workflow
     * [openmc_tally_unit_converter](https://github.com/openmc-data-storage/openmc_tally_unit_converter) converts the units of common tally such as
     heating, DPA, effective dose into user specified units. 
 
-    * [openmc_data_downloader](https://github.com/openmc-data-storage/openmc_data_downloader) performs on the fly downloading of nuclear data
-    needed for OpenMC neutronics simulations.
+    * [openmc_data_downloader](https://github.com/openmc-data-storage/openmc_data_downloader) performs on the fly downloading of nuclear data needed for OpenMC neutronics simulations.
+
+    * [openmc_data](https://github.com/openmc-data-storage/openmc_data) downloading and processing of nuclear data needed for OpenMC neutronics simulations.
 
     * [openmc-plasma-source](https://github.com/fusion-energy/openmc-plasma-source/)
     Creates a plasma source as an openmc.source object from input parameters that describe the plasma 

diff --git a/environment.yml b/environment.yml
@@ -0,0 +1,16 @@
+name: base
+channels:
+  - fusion-energy
+  - cadquery
+  - conda-forge
+dependencies:
+  - paramak>=0.8.1
+  - pip
+  - pip:
+    - openmc-dagmc-wrapper
+    - openmc-plasma-source
+    - openmc_tally_unit_converter
+    - spectrum_plotter
+    - regular_mesh_plotter
+    - openmc_data_downloader
+    - openmc_data
diff --git a/example_01_single_volume_cell_tally/2_run_neutronics_simulation.py b/example_01_single_volume_cell_tally/2_run_neutronics_simulation.py
@@ -10,35 +10,26 @@
 
 
 # defines a simple DT neutron point source
-my_source = ops.FusionPointSource(
-    coordinate = (0,0,0),
-    temperature = 20000.,
-    fuel='DT'
-)
+my_source = ops.FusionPointSource(coordinate=(0, 0, 0), temperature=20000.0, fuel="DT")
 
 # set the geometry file to use
 geometry = odw.Geometry(
-    h5m_filename='dagmc.h5m',
+    h5m_filename="dagmc.h5m",
 )
 
 # sets the material to use
 materials = odw.Materials(
-    h5m_filename='dagmc.h5m',
-    correspondence_dict={"mat_my_material": "eurofer"}
+    h5m_filename="dagmc.h5m", correspondence_dict={"my_material": "eurofer"}
 )
 
 # creates a cell tally for neutron flux
 tally1 = odw.CellTally(
-    tally_type="neutron_flux",
-    target="mat_my_material",
-    materials=materials
+    tally_type="neutron_flux", target="my_material", materials=materials
 )
 
 # creates a cell tally for neutron spectra
 tally2 = odw.CellTally(
-    tally_type="neutron_spectra",
-    target="mat_my_material",
-    materials=materials
+    tally_type="neutron_spectra", target="my_material", materials=materials
 )
 
 tallies = openmc.Tallies([tally1, tally2])
@@ -63,7 +54,7 @@
 statepoint = openmc.StatePoint(filepath=statepoint_file)
 
 # gets the first tally using its name
-my_tally_1 = statepoint.get_tally(name="mat_my_material_neutron_flux")
+my_tally_1 = statepoint.get_tally(name="my_material_neutron_flux")
 
 # gets number of neutron for a 1.3 mega joule shot
 source_strength = otuc.find_source_strength(fusion_energy_per_second_or_per_pulse=1.3e6)
@@ -78,25 +69,25 @@
 print(f"flux per second = {result}", end="\n\n")
 
 # gets the second tally using its name
-my_tally_2 = statepoint.get_tally(name="mat_my_material_neutron_spectra")
+my_tally_2 = statepoint.get_tally(name="my_material_neutron_spectra")
 
 # returns the tally converted to MeV, scaled and normalisation for source strength
 result = otuc.process_spectra_tally(
     tally=my_tally_2,
     required_units="centimeter / second",
-    required_energy_units='MeV',
-    source_strength=source_strength
+    required_energy_units="MeV",
+    source_strength=source_strength,
 )
 
 # creates a matplotlib figure of the tally
 spectrum_plot = plot_spectrum_from_tally(
-    spectrum={'neutron spectra tally': my_tally_2},
+    spectrum={"neutron spectra tally": my_tally_2},
     x_label="Energy [MeV]",
     y_label="Flux [n/cm^2s]",
     x_scale="log",
     y_scale="log",
     title="neutron spectra tally",
-    filename='my_spectrum_plot.png'
+    filename="my_spectrum_plot.png",
 )
 
-print('image saved as my_spectrum_plot.png')
+print("image saved as my_spectrum_plot.png")
diff --git a/example_02_multi_volume_cell_tally/2_run_neutronics_simulation.py b/example_02_multi_volume_cell_tally/2_run_neutronics_simulation.py
@@ -20,14 +20,14 @@
 materials = odw.Materials(
     h5m_filename=geometry.h5m_filename,
     correspondence_dict={
-        "mat_plasma": "DT_plasma",
-        "mat_inboard_tf_coils": "copper",
-        "mat_center_column_shield": "tungsten",
-        "mat_firstwall": "tungsten",
-        "mat_blanket": "Li4SiO4",
-        "mat_blanket_rear_wall": "eurofer",
-        "mat_divertor_upper": "tungsten",
-        "mat_divertor_lower": "tungsten",
+        "plasma": "DT_plasma",
+        "inboard_tf_coils": "copper",
+        "center_column_shield": "tungsten",
+        "firstwall": "tungsten",
+        "blanket": "Li4SiO4",
+        "blanket_rear_wall": "eurofer",
+        "divertor_upper": "tungsten",
+        "divertor_lower": "tungsten",
     }
 )
 

diff --git a/example_04_multi_volume_regular_mesh_tally/2_run_neutronics_simulation.py b/example_04_multi_volume_regular_mesh_tally/2_run_neutronics_simulation.py
@@ -13,13 +13,13 @@
 # these materials are input as strings so they will be looked up in the
 # neutronics material maker package
 material_tag_to_material_dict = {
-    "mat_inboard_tf_coils": "copper",
-    "mat_center_column_shield": "tungsten",
-    "mat_firstwall": "tungsten",
-    "mat_blanket": "Li4SiO4",
-    "mat_blanket_rear_wall": "eurofer",
-    "mat_divertor_upper": "tungsten",
-    "mat_divertor_lower": "tungsten",
+    "inboard_tf_coils": "copper",
+    "center_column_shield": "tungsten",
+    "firstwall": "tungsten",
+    "blanket": "Li4SiO4",
+    "blanket_rear_wall": "eurofer",
+    "divertor_upper": "tungsten",
+    "divertor_lower": "tungsten",
 }
 
 materials = odw.Materials(