fixing nuc data for task 9

tasks/task_09_CSG_instantaneous_dose_tallies/1_surface_dose_from_gamma_source.ipynb

@@ -27,6 +27,26 @@
     "The following section plots effective dose coefficient as a function of incident particle energy for neutrons and photons."
    ]
   },
+  {
+   "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 plotly.graph_objects as go\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": "code",
    "execution_count": null,
@@ -35,9 +55,6 @@
    },
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "import plotly.graph_objects as go\n",
-    "\n",
     "energy_bins_n, dose_coeffs_n = openmc.data.dose_coefficients(\n",
     "    particle='neutron',\n",
     "    geometry='AP'  # AP defines the direction of the source to person, for more details see documentation https://docs.openmc.org/en/stable/pythonapi/generated/openmc.data.dose_coefficients.html\n",

tasks/task_09_CSG_instantaneous_dose_tallies/2_surface_dose_from_gamma_source_study.ipynb

@@ -22,6 +22,26 @@
     "This is the similar to the previous task geometry, but this time we have nested spheres which will be used to tally the dose on."
    ]
   },
+  {
+   "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": "code",
    "execution_count": null,

tasks/task_09_CSG_instantaneous_dose_tallies/3_cell_dose_from_neutron.py

@@ -7,6 +7,11 @@
 
 
 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'
+
 import math
 import matplotlib.pyplot as plt
 

tasks/task_09_CSG_instantaneous_dose_tallies/4_cell_dose_from_photon.py

@@ -12,6 +12,10 @@
 import matplotlib.pyplot as plt
 
 
+# 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'
+
 # Tissue Equivalent, MS20 from PNNL
 mat_tissue = openmc.Material()
 mat_tissue.add_element("O", 0.079013)

tasks/task_09_CSG_instantaneous_dose_tallies/5_mesh_dose_from_neutrons.py

@@ -3,6 +3,9 @@
 import openmc
 import matplotlib.pyplot as plt
 
+# 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'
 
 mat = openmc.Material()
 mat.add_element("Al", 1)

tasks/task_09_CSG_instantaneous_dose_tallies/compare_dose_simulation_with_back_of_envelope.py

@@ -10,6 +10,9 @@
 import matplotlib.pyplot as plt
 import numpy as np
 
+# 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'
 
 def manual_dose_calc(
     particles_per_shot:int,