fixing task 5 nuc data

tasks/task_05_CSG_cell_tally_TBR/1_example_tritium_production.ipynb

@@ -23,9 +23,14 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This example uses a simple sphere model with a breeder material and a tritium production tally.\n",
-    "\n",
-    "This first code block makes the geometry, materials and settings for the neutronics model."
+    "This example uses a simple sphere model with a breeder material and a tritium production tally."
+   ]
+  },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -36,6 +41,24 @@
    "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": [
+    "This first code block makes the geometry, materials and settings for the neutronics model."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "breeder_material = openmc.Material()  # lithium lead chemical equation is Pb84.2Li15.8\n",

tasks/task_05_CSG_cell_tally_TBR/2_example_tritium_production_study.ipynb

@@ -11,6 +11,18 @@
     "This python notebook allows users to make a simple model with a Lithium blanket of controllable enrichment. Simulations are then performed for several different Li6 enrichments and the TBR found."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": []
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": []
+  },
   {
    "cell_type": "markdown",
    "metadata": {},
@@ -24,9 +36,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
-    "\n",
     "def make_materials_geometry_tallies(enrichment):\n",
     "    \"\"\"Makes a neutronics model of a blanket and simulates the TBR value.\n",
     "\n",

tasks/task_05_CSG_cell_tally_TBR/3_example_tritium_production_study_with_openmc_lib.py

@@ -14,6 +14,9 @@
 
 import openmc
 
+# Setting the cross section path to the correct location in the docker image.
+# If you are running this outside the docker image you will have to change this path to your local cross section path.
+openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'
 
 # make some python materials
 breeder_material = openmc.Material(material_id = 12)  # Pb84.2Li15.8