Merge pull request #295 from fusion-energy/remove_nuc_data_env_from_d…

…ocker

Remove nuc data env from docker

.devcontainer/base.Dockerfile

@@ -118,7 +118,7 @@ RUN pip install neutronics_material_maker[density] \
 RUN pip install git+https://github.com/fusion-energy/openmc-plasma-source
 
 RUN pip install git+https://github.com/CadQuery/cadquery.git@bc82cb04c59668a1369d9ce648361c8786bbd1c8 --no-deps
-RUN pip install cadquery-ocp==7.7.1 "multimethod>=1.7<2.0" nlopt typish casadi path ezdxf nptyping==2.0.1
+RUN pip install cadquery-ocp==7.7.1 "multimethod>=1.7,<2.0" nlopt typish casadi path ezdxf nptyping==2.0.1
 
 # Python libraries used in the workshop
 RUN pip install cmake\
@@ -170,7 +170,7 @@ RUN if [ "$build_double_down" = "ON" ] ; \
 RUN mkdir MOAB && \
     cd MOAB && \
     # newer versions of moab (5.4.0, 5.4.1) don't produce an importable pymoab package!
-    # TODO try moab 5.5.0
+    # TODO try moab 5.5.0 and 5.5.1
     git clone  --single-branch --branch 5.3.1 --depth 1 https://bitbucket.org/fathomteam/moab.git && \
     mkdir build && \
     cd build && \
@@ -249,6 +249,3 @@ RUN download_endf_chain -d nuclear_data -r b8.0
 RUN wget https://github.com/mit-crpg/WMP_Library/releases/download/v1.1/WMP_Library_v1.1.tar.gz && \
     tar -xf WMP_Library_v1.1.tar.gz -C /  && \
     rm WMP_Library_v1.1.tar.gz
-
-ENV OPENMC_CROSS_SECTIONS=/nuclear_data/cross_sections.xml
-ENV OPENMC_CHAIN_FILE=/nuclear_data/chain-endf-b8.0.xml

tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb

@@ -27,6 +27,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/KYIsDjip1nQ\", width=560, height=340)"
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import packages needed and configure the OpenMC nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import matplotlib.pyplot as plt\n",
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "markdown",
    "metadata": {},
@@ -42,9 +62,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "import matplotlib.pyplot as plt\n",
-    "\n",
     "# MATERIALS\n",
     "\n",
     "# creates two materials, one is a neutron multiplier (lead) and the other a tritium breeder (lithium)\n",

tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb

@@ -20,6 +20,25 @@
     "This first code block defines the model geometry, materials and neutron source."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": 1,
@@ -28,8 +47,6 @@
    },
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
     "# MATERIALS\n",
     "\n",
     "breeder_material = openmc.Material()   # Pb84.2Li15.8\n",

tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb

@@ -18,6 +18,25 @@
     "This first code block defines the model geometry and materials."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,