Adding nuc data path to all

.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_01_cross_sections/1_isotope_xs_plot.ipynb

@@ -29,6 +29,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/eBZ2lY_2v7IA\", width=560, height=340)"
    ]
   },
+  {
+   "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",
+    "\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": {},
@@ -50,7 +70,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
     "import matplotlib.pyplot as plt\n",
     "\n",
     "fig = openmc.plotter.plot_xs(\n",

tasks/task_01_cross_sections/2_element_xs_plot.ipynb

@@ -23,6 +23,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/ELZNeIdSuMY\", width=560, height=340)"
    ]
   },
+  {
+   "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",
+    "\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": {},

tasks/task_01_cross_sections/3_material_xs_plot.ipynb

@@ -37,7 +37,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This code block imports a function and packages needed for the task."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -46,7 +46,11 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc"
+    "import openmc\n",
+    "\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'"
    ]
   },
   {

tasks/task_01_cross_sections/4_Doppler_broadening.ipynb

@@ -22,16 +22,28 @@
     "IFrame(src=\"https://www.youtube.com/embed/mkl1mVnTO6g\", width=560, height=340)"
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import packages needed for OpenMC and configure the nuclear data path"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
    "metadata": {},
    "outputs": [],
    "source": [
     "import numpy as np\n",
+    "import matplotlib.pyplot as plt\n",
+    "\n",
     "import openmc\n",
     "from openmc.data import REACTION_MT\n",
-    "import matplotlib.pyplot as plt"
+    "\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'"
    ]
   },
   {

tasks/task_02_making_materials/1_example_materials_from_isotopes.ipynb

@@ -23,6 +23,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/-NGnY-1TWCA\", width=560, height=340)"
    ]
   },
+  {
+   "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",
+    "\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": {},

tasks/task_02_making_materials/2_example_materials_from_elements.ipynb

@@ -15,7 +15,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "The following code block is a simple example of creating a material (water H2O) from elements. (Note how Hydrogen and Oxygen elements have been specified rather than each specific isotope)."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -26,6 +26,24 @@
    "source": [
     "import openmc\n",
     "\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": {},
+   "source": [
+    "The following code block is a simple example of creating a material (water H2O) from elements. (Note how Hydrogen and Oxygen elements have been specified rather than each specific isotope)."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# Making water from elements\n",
     "\n",
     "water_mat = openmc.Material()\n",

tasks/task_02_making_materials/5_example_materials_mixed.ipynb

@@ -24,15 +24,25 @@
     "This python notebook allows users to make mixed materials using OpenMC and the Neutronics Material Maker."
    ]
   },
+  {
+   "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 the packages needed for the task\n",
+    "import neutronics_material_maker as nmm\n",
     "import openmc\n",
-    "import neutronics_material_maker as nmm"
+    "\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'"
    ]
   },
   {

tasks/task_03_making_CSG_geometry/1_simple_csg_geometry.ipynb

@@ -22,6 +22,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/Ovr7oYukYRw\", width=560, height=340)"
    ]
   },
+  {
+   "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",
+    "\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'"
+   ]
+  },
   {
    "attachments": {},
    "cell_type": "markdown",
@@ -38,8 +58,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
     "# example surfaces\n",
     "inner_sphere_surface = openmc.Sphere(r=500)\n",
     "outer_sphere_surface = openmc.Sphere(r=600)\n",

tasks/task_03_making_CSG_geometry/2_intermediate_csg_geometry.ipynb

@@ -9,8 +9,33 @@
     "\n",
     "This python notebook allows users to build upon the geometry constructed in Part 1 by adding a center column to the model.\n",
     "\n",
-    "The center column requires a different surface type. OpenMC has several different surface types in addition to spheres. The documentation contains useful information on all surface types supported: https://openmc.readthedocs.io/en/stable/usersguide/geometry.html#surfaces-and-regions.\n",
+    "The center column requires a different surface type. OpenMC has several different surface types in addition to spheres. The documentation contains useful information on all surface types supported: https://openmc.readthedocs.io/en/stable/usersguide/geometry.html#surfaces-and-regions."
+   ]
+  },
+  {
+   "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",
     "\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": {},
+   "source": [
     "For the center column we will use a cylinder with Z axis orientation.\n",
     "\n",
     "The code block below shows how the center column can be defined."
@@ -22,8 +47,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
     "# surfaces\n",
     "central_column_surface = openmc.ZCylinder(r=100) # note the new surface type\n",
     "inner_sphere_surface = openmc.Sphere(r=480)\n",

tasks/task_03_making_CSG_geometry/3_viewing_the_geometry_as_vtk.ipynb

@@ -24,6 +24,13 @@
     "This first code block recreates the simple reactor geometry seen in Part 2, but also assigns materials to each cell."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
@@ -32,6 +39,17 @@
    "source": [
     "import openmc\n",
     "\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,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "copper = openmc.Material()\n",
     "copper.set_density('g/cm3', 8.5)\n",
     "copper.add_element('Cu', 1.0)  # Note, percent_type does not have to be specified as material is 100% copper\n",

tasks/task_04_make_sources/1_point_source_plots.ipynb

@@ -22,6 +22,13 @@
     "IFrame(src=\"https://www.youtube.com/embed/j9dT1Viqcu4\", width=560, height=340)"
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import packages needed"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,

tasks/task_04_make_sources/2_ring_source.ipynb

@@ -38,6 +38,13 @@
     "In this example we are going to make use of the Uniform disitrbution by feel free to tinker with the differnt distributions an make your own source."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import packages needed"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
@@ -46,11 +53,11 @@
    "source": [
     "import openmc\n",
     "\n",
-    "# provides simple source plotting functions\n",
+    "\n",
+    "# This package provides simple source plotting functions\n",
     "# for more details here is a link to the package repository\n",
     "# https://github.com/fusion-energy/openmc_source_plotter\n",
-    "from openmc_source_plotter import plot_source_position\n",
-    "\n"
+    "from openmc_source_plotter import plot_source_position"
    ]
   },
   {
@@ -148,7 +155,7 @@
    "name": "python",
    "nbconvert_exporter": "python",
    "pygments_lexer": "ipython3",
-   "version": "3.10.12"
+   "version": "3.12.2"
   }
  },
  "nbformat": 4,