diff --git a/tasks/task_01_cross_sections/1_isotope_xs_plot.ipynb b/tasks/task_01_cross_sections/1_isotope_xs_plot.ipynb
index 70c8c40b..bacf4c23 100644
--- a/tasks/task_01_cross_sections/1_isotope_xs_plot.ipynb
+++ b/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",
diff --git a/tasks/task_01_cross_sections/2_element_xs_plot.ipynb b/tasks/task_01_cross_sections/2_element_xs_plot.ipynb
index 739f6832..f0e8683a 100644
--- a/tasks/task_01_cross_sections/2_element_xs_plot.ipynb
+++ b/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": {},
diff --git a/tasks/task_01_cross_sections/3_material_xs_plot.ipynb b/tasks/task_01_cross_sections/3_material_xs_plot.ipynb
index 4b510038..832ccdff 100644
--- a/tasks/task_01_cross_sections/3_material_xs_plot.ipynb
+++ b/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'"
    ]
   },
   {
diff --git a/tasks/task_01_cross_sections/4_Doppler_broadening.ipynb b/tasks/task_01_cross_sections/4_Doppler_broadening.ipynb
index ac7a9a2c..d4bd5d9a 100644
--- a/tasks/task_01_cross_sections/4_Doppler_broadening.ipynb
+++ b/tasks/task_01_cross_sections/4_Doppler_broadening.ipynb
@@ -22,6 +22,13 @@
     "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,
@@ -29,9 +36,14 @@
    "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'"
    ]
   },
   {
diff --git a/tasks/task_02_making_materials/1_example_materials_from_isotopes.ipynb b/tasks/task_02_making_materials/1_example_materials_from_isotopes.ipynb
index 199725c9..d80250c1 100644
--- a/tasks/task_02_making_materials/1_example_materials_from_isotopes.ipynb
+++ b/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": {},
diff --git a/tasks/task_02_making_materials/2_example_materials_from_elements.ipynb b/tasks/task_02_making_materials/2_example_materials_from_elements.ipynb
index 9c318473..abb95c83 100644
--- a/tasks/task_02_making_materials/2_example_materials_from_elements.ipynb
+++ b/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",
diff --git a/tasks/task_02_making_materials/5_example_materials_mixed.ipynb b/tasks/task_02_making_materials/5_example_materials_mixed.ipynb
index 9a3a0c04..add3af91 100644
--- a/tasks/task_02_making_materials/5_example_materials_mixed.ipynb
+++ b/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'"
    ]
   },
   {
diff --git a/tasks/task_03_making_CSG_geometry/1_simple_csg_geometry.ipynb b/tasks/task_03_making_CSG_geometry/1_simple_csg_geometry.ipynb
index bc7e31e9..29e9b97a 100644
--- a/tasks/task_03_making_CSG_geometry/1_simple_csg_geometry.ipynb
+++ b/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",
diff --git a/tasks/task_03_making_CSG_geometry/2_intermediate_csg_geometry.ipynb b/tasks/task_03_making_CSG_geometry/2_intermediate_csg_geometry.ipynb
index a7dd5589..995c8c2f 100644
--- a/tasks/task_03_making_CSG_geometry/2_intermediate_csg_geometry.ipynb
+++ b/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",
diff --git a/tasks/task_03_making_CSG_geometry/3_viewing_the_geometry_as_vtk.ipynb b/tasks/task_03_making_CSG_geometry/3_viewing_the_geometry_as_vtk.ipynb
index 644e1e0b..2af9dcf4 100644
--- a/tasks/task_03_making_CSG_geometry/3_viewing_the_geometry_as_vtk.ipynb
+++ b/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",
diff --git a/tasks/task_04_make_sources/1_point_source_plots.ipynb b/tasks/task_04_make_sources/1_point_source_plots.ipynb
index 72f91cfc..70a6173b 100644
--- a/tasks/task_04_make_sources/1_point_source_plots.ipynb
+++ b/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,
diff --git a/tasks/task_04_make_sources/2_ring_source.ipynb b/tasks/task_04_make_sources/2_ring_source.ipynb
index bd9cf3aa..726e4f7d 100644
--- a/tasks/task_04_make_sources/2_ring_source.ipynb
+++ b/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,
diff --git a/tasks/task_04_make_sources/5_gamma_source_example.ipynb b/tasks/task_04_make_sources/5_gamma_source_example.ipynb
index d2b6ff7f..f20071e7 100644
--- a/tasks/task_04_make_sources/5_gamma_source_example.ipynb
+++ b/tasks/task_04_make_sources/5_gamma_source_example.ipynb
@@ -96,7 +96,7 @@
    "outputs": [],
    "source": [
     "# remove any other xml files so that openmc reads the correct xml files\n",
-    "!rm *.xml\n",
+    "!rm *.xml || true\n",
     "\n",
     "# number_of_particles can be increased to sample more particles and the histogram resoluton can be changed\n",
     "plot_source_energy(\n",
diff --git a/tasks/task_04_make_sources/7_strucutured_mesh_source.py b/tasks/task_04_make_sources/7_strucutured_mesh_source.py
index aa4828df..79c6dda5 100644
--- a/tasks/task_04_make_sources/7_strucutured_mesh_source.py
+++ b/tasks/task_04_make_sources/7_strucutured_mesh_source.py
@@ -6,7 +6,8 @@
 import openmc
 import numpy as np
 
-# setting the nuclear data path to the correct location in the docker image
+# Setting the nuclear data 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.
 openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'
 
 # making a minimal geometry
diff --git a/tasks/task_05_CSG_cell_tally_TBR/1_example_tritium_production.ipynb b/tasks/task_05_CSG_cell_tally_TBR/1_example_tritium_production.ipynb
index 8dd85253..a9b089a9 100644
--- a/tasks/task_05_CSG_cell_tally_TBR/1_example_tritium_production.ipynb
+++ b/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",
diff --git a/tasks/task_05_CSG_cell_tally_TBR/2_example_tritium_production_study.ipynb b/tasks/task_05_CSG_cell_tally_TBR/2_example_tritium_production_study.ipynb
index e0a46b93..af3c2d59 100644
--- a/tasks/task_05_CSG_cell_tally_TBR/2_example_tritium_production_study.ipynb
+++ b/tasks/task_05_CSG_cell_tally_TBR/2_example_tritium_production_study.ipynb
@@ -15,7 +15,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This first code block a neutronics model for the provided enrichment value, runs the simulation and returns the TBR value obtained."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -26,7 +26,24 @@
    "source": [
     "import openmc\n",
     "\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 a neutronics model for the provided enrichment value, runs the simulation and returns the TBR value obtained."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "def make_materials_geometry_tallies(enrichment):\n",
     "    \"\"\"Makes a neutronics model of a blanket and simulates the TBR value.\n",
     "\n",
diff --git a/tasks/task_05_CSG_cell_tally_TBR/3_example_tritium_production_study_with_openmc_lib.py b/tasks/task_05_CSG_cell_tally_TBR/3_example_tritium_production_study_with_openmc_lib.py
index c8cd8ecd..fca79b09 100644
--- a/tasks/task_05_CSG_cell_tally_TBR/3_example_tritium_production_study_with_openmc_lib.py
+++ b/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
diff --git a/tasks/task_06_CSG_cell_tally_DPA/1_find_dpa.ipynb b/tasks/task_06_CSG_cell_tally_DPA/1_find_dpa.ipynb
index f3d0c963..4c52fd7d 100644
--- a/tasks/task_06_CSG_cell_tally_DPA/1_find_dpa.ipynb
+++ b/tasks/task_06_CSG_cell_tally_DPA/1_find_dpa.ipynb
@@ -27,7 +27,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This first stage sets up the geometry and materials for the simulation."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -38,7 +38,24 @@
    "source": [
     "import openmc\n",
     "\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 stage sets up the geometry and materials for the simulation."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "density_of_iron_in_g_per_cm3 = 7.75\n",
diff --git a/tasks/task_07_CSG_cell_tally_spectra/1_example_neutron_flux_and_units.ipynb b/tasks/task_07_CSG_cell_tally_spectra/1_example_neutron_flux_and_units.ipynb
index 897a4280..e5ae6ddc 100644
--- a/tasks/task_07_CSG_cell_tally_spectra/1_example_neutron_flux_and_units.ipynb
+++ b/tasks/task_07_CSG_cell_tally_spectra/1_example_neutron_flux_and_units.ipynb
@@ -13,7 +13,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -24,6 +24,24 @@
    "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 section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "# Due to the hydrogen content water is a very good neutron moderator\n",
@@ -115,7 +133,7 @@
     "model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)\n",
     "\n",
     "# remove old files and runs OpenMC\n",
-    "!rm *.h5\n",
+    "!rm *.h5 || true\n",
     "results_filename = model.run()"
    ]
   },
diff --git a/tasks/task_07_CSG_cell_tally_spectra/2_example_neutron_spectra_on_cell.ipynb b/tasks/task_07_CSG_cell_tally_spectra/2_example_neutron_spectra_on_cell.ipynb
index 92268920..4401f117 100644
--- a/tasks/task_07_CSG_cell_tally_spectra/2_example_neutron_spectra_on_cell.ipynb
+++ b/tasks/task_07_CSG_cell_tally_spectra/2_example_neutron_spectra_on_cell.ipynb
@@ -25,7 +25,9 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+    "\n",
+    "\n",
+    "First import OpenMC and configure the nuclear data path\n"
    ]
   },
   {
@@ -34,10 +36,28 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
     "import numpy as np\n",
     "import matplotlib.pyplot as plt\n",
+    "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 section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "# Due to the hydrogen content water is a very good neutron moderator\n",
@@ -125,7 +145,7 @@
     "model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)\n",
     "\n",
     "# remove old files and runs OpenMC\n",
-    "!rm *.h5\n",
+    "!rm *.h5 || true\n",
     "results_filename = model.run()"
    ]
   },
diff --git a/tasks/task_07_CSG_cell_tally_spectra/3_example_neutron_spectra_on_surface.ipynb b/tasks/task_07_CSG_cell_tally_spectra/3_example_neutron_spectra_on_surface.ipynb
index 71b98653..5cb18dc2 100644
--- a/tasks/task_07_CSG_cell_tally_spectra/3_example_neutron_spectra_on_surface.ipynb
+++ b/tasks/task_07_CSG_cell_tally_spectra/3_example_neutron_spectra_on_surface.ipynb
@@ -29,7 +29,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -40,6 +40,24 @@
    "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 section creates a simple material, geometry and settings. This model is used in both the neutron current tally and the neutron flux tally."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "# Due to the hydrogen content water is a very good neutron moderator\n",
@@ -158,7 +176,7 @@
     "model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)\n",
     "\n",
     "# remove old files and runs OpenMC\n",
-    "!rm *.h5\n",
+    "!rm *.h5 || true\n",
     "results_filename = model.run()"
    ]
   },
diff --git a/tasks/task_07_CSG_cell_tally_spectra/4_example_photon_spectra.ipynb b/tasks/task_07_CSG_cell_tally_spectra/4_example_photon_spectra.ipynb
index 9c9c1313..7315ec5a 100644
--- a/tasks/task_07_CSG_cell_tally_spectra/4_example_photon_spectra.ipynb
+++ b/tasks/task_07_CSG_cell_tally_spectra/4_example_photon_spectra.ipynb
@@ -19,7 +19,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This section creates a simple material, geometry and settings. This model is used in both the photon current tally and photon flux tally."
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -28,9 +28,27 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
     "import matplotlib.pyplot as plt\n",
+    "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 section creates a simple material, geometry and settings. This model is used in both the photon current tally and photon flux tally."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "# Tungsten is a very good photon shield, partly due to its high Z number and electrons\n",
@@ -122,7 +140,7 @@
     "model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)\n",
     "\n",
     "# remove old files and runs OpenMC\n",
-    "!rm *.h5\n",
+    "!rm *.h5 || true\n",
     "results_filename = model.run()"
    ]
   },
diff --git a/tasks/task_07_CSG_cell_tally_spectra/5_neutron_energy_loss_plot.ipynb b/tasks/task_07_CSG_cell_tally_spectra/5_neutron_energy_loss_plot.ipynb
index 58a7cf01..7693c2f4 100644
--- a/tasks/task_07_CSG_cell_tally_spectra/5_neutron_energy_loss_plot.ipynb
+++ b/tasks/task_07_CSG_cell_tally_spectra/5_neutron_energy_loss_plot.ipynb
@@ -19,7 +19,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "Imports required packages"
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -28,9 +28,14 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
     "import matplotlib.pyplot as plt\n",
-    "from shutil import move"
+    "from shutil import move\n",
+    "\n",
+    "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'"
    ]
   },
   {
@@ -112,7 +117,7 @@
    "outputs": [],
    "source": [
     "# remove old summary and statepoint files\n",
-    "!rm *.h5\n",
+    "!rm *.h5 || true\n",
     "\n",
     "for element, color in zip([liquid_hydrogen, tungsten], [\"red\", \"blue\"]):\n",
     "    plt.plot([0], label=element, color=color)\n",
diff --git a/tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb
index 82006429..8d156757 100644
--- a/tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb
+++ b/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",
diff --git a/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb
index f47175cd..c3c25fa5 100644
--- a/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb
+++ b/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb
@@ -20,16 +20,33 @@
     "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,
    "metadata": {
     "scrolled": true
    },
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
     "# MATERIALS\n",
     "\n",
     "breeder_material = openmc.Material()   # Pb84.2Li15.8\n",
@@ -110,7 +127,7 @@
   },
   {
    "cell_type": "code",
-   "execution_count": null,
+   "execution_count": 2,
    "metadata": {},
    "outputs": [],
    "source": [
@@ -131,7 +148,7 @@
   },
   {
    "cell_type": "code",
-   "execution_count": null,
+   "execution_count": 3,
    "metadata": {},
    "outputs": [],
    "source": [
@@ -162,15 +179,100 @@
   },
   {
    "cell_type": "code",
-   "execution_count": null,
+   "execution_count": 6,
    "metadata": {},
-   "outputs": [],
+   "outputs": [
+    {
+     "name": "stdout",
+     "output_type": "stream",
+     "text": [
+      "rm: cannot remove 's*.h5': No such file or directory\n",
+      "                                %%%%%%%%%%%%%%%\n",
+      "                           %%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                        %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                      %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                    %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                   %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                                    %%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                                     %%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                 ###############      %%%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                ##################     %%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                ###################     %%%%%%%%%%%%%%%%%%%%%%%\n",
+      "                ####################     %%%%%%%%%%%%%%%%%%%%%%\n",
+      "                #####################     %%%%%%%%%%%%%%%%%%%%%\n",
+      "                ######################     %%%%%%%%%%%%%%%%%%%%\n",
+      "                #######################     %%%%%%%%%%%%%%%%%%\n",
+      "                 #######################     %%%%%%%%%%%%%%%%%\n",
+      "                 ######################     %%%%%%%%%%%%%%%%%\n",
+      "                  ####################     %%%%%%%%%%%%%%%%%\n",
+      "                    #################     %%%%%%%%%%%%%%%%%\n",
+      "                     ###############     %%%%%%%%%%%%%%%%\n",
+      "                       ############     %%%%%%%%%%%%%%%\n",
+      "                          ########     %%%%%%%%%%%%%%\n",
+      "                                      %%%%%%%%%%%\n",
+      "\n",
+      "                 | The OpenMC Monte Carlo Code\n",
+      "       Copyright | 2011-2024 MIT, UChicago Argonne LLC, and contributors\n",
+      "         License | https://docs.openmc.org/en/latest/license.html\n",
+      "         Version | 0.15.1-dev\n",
+      "        Git SHA1 | 86fc40a6456e789676eec3410afff38e340ea8f8\n",
+      "       Date/Time | 2024-08-30 11:57:21\n",
+      "  OpenMP Threads | 16\n",
+      "\n",
+      " Reading model XML file 'model.xml' ...\n",
+      " Reading cross sections XML file...\n",
+      " Reading Pb204 from /nuclear_data/endfb-viii.0-hdf5/neutron/Pb204.h5\n",
+      " Reading Pb206 from /nuclear_data/endfb-viii.0-hdf5/neutron/Pb206.h5\n",
+      " Reading Pb207 from /nuclear_data/endfb-viii.0-hdf5/neutron/Pb207.h5\n",
+      " Reading Pb208 from /nuclear_data/endfb-viii.0-hdf5/neutron/Pb208.h5\n",
+      " Reading Li6 from /nuclear_data/endfb-viii.0-hdf5/neutron/Li6.h5\n",
+      " Reading Li7 from /nuclear_data/endfb-viii.0-hdf5/neutron/Li7.h5\n",
+      " Reading Cu63 from /nuclear_data/endfb-viii.0-hdf5/neutron/Cu63.h5\n",
+      " Reading Cu65 from /nuclear_data/endfb-viii.0-hdf5/neutron/Cu65.h5\n",
+      " Reading Fe54 from /nuclear_data/endfb-viii.0-hdf5/neutron/Fe54.h5\n",
+      " Reading Fe56 from /nuclear_data/endfb-viii.0-hdf5/neutron/Fe56.h5\n",
+      " Reading Fe57 from /nuclear_data/endfb-viii.0-hdf5/neutron/Fe57.h5\n",
+      " Reading Fe58 from /nuclear_data/endfb-viii.0-hdf5/neutron/Fe58.h5\n",
+      " Minimum neutron data temperature: 294 K\n",
+      " Maximum neutron data temperature: 294 K\n",
+      " Preparing distributed cell instances...\n",
+      " Writing summary.h5 file...\n",
+      " Maximum neutron transport energy: 20000000 eV for Li6\n",
+      "\n",
+      " ===============>     FIXED SOURCE TRANSPORT SIMULATION     <===============\n",
+      "\n",
+      " Simulating batch 1\n",
+      " Simulating batch 2\n",
+      " Creating state point statepoint.2.h5...\n",
+      "\n",
+      " =======================>     TIMING STATISTICS     <=======================\n",
+      "\n",
+      " Total time for initialization     = 1.1259e+00 seconds\n",
+      "   Reading cross sections          = 1.1053e+00 seconds\n",
+      " Total time in simulation          = 4.0207e-01 seconds\n",
+      "   Time in transport only          = 3.9833e-01 seconds\n",
+      "   Time in active batches          = 4.0207e-01 seconds\n",
+      "   Time accumulating tallies       = 3.6221e-04 seconds\n",
+      "   Time writing statepoints        = 3.2080e-03 seconds\n",
+      " Total time for finalization       = 1.3965e-02 seconds\n",
+      " Total time elapsed                = 1.5450e+00 seconds\n",
+      " Calculation Rate (active)         = 24871.5 particles/second\n",
+      "\n",
+      " ============================>     RESULTS     <============================\n",
+      "\n",
+      " Leakage Fraction            = 0.14289 +/- 0.00061\n",
+      "\n"
+     ]
+    }
+   ],
    "source": [
     "# deletes old statepoint and summary files\n",
     "!rm s*.h5\n",
     "\n",
     "# Run OpenMC!\n",
     "model = openmc.Model(my_geometry, mats, sett, tallies)\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'\n",
+    "\n",
     "sp_filename = model.run()"
    ]
   },
@@ -184,9 +286,20 @@
   },
   {
    "cell_type": "code",
-   "execution_count": null,
+   "execution_count": 7,
    "metadata": {},
-   "outputs": [],
+   "outputs": [
+    {
+     "data": {
+      "text/plain": [
+       "<vtkmodules.vtkCommonDataModel.vtkStructuredGrid(0x5a14f1d74570) at 0x73290807fa60>"
+      ]
+     },
+     "execution_count": 7,
+     "metadata": {},
+     "output_type": "execute_result"
+    }
+   ],
    "source": [
     "\n",
     "# loads up the output file from the simulation\n",
@@ -225,6 +338,37 @@
     "Visit can be downloaded here: https://wci.llnl.gov/simulation/computer-codes/visit/downloads."
    ]
   },
+  {
+   "cell_type": "code",
+   "execution_count": 90,
+   "metadata": {},
+   "outputs": [
+    {
+     "data": {
+      "application/vnd.jupyter.widget-view+json": {
+       "model_id": "6de27a09e3594a6bbb7cb8846e9781ae",
+       "version_major": 2,
+       "version_minor": 0
+      },
+      "text/plain": [
+       "Widget(value='<iframe src=\"http://localhost:35203/index.html?ui=P_0x73263d9a6b60_52&reconnect=auto\" class=\"pyv…"
+      ]
+     },
+     "metadata": {},
+     "output_type": "display_data"
+    }
+   ],
+   "source": [
+    "import pyvista as pv\n",
+    "mesh = pv.read('heating_tally_on_reg_mesh.vtk')\n",
+    "# mesh.plot()\n",
+    "mesh.set_active_scalars(name='mean')\n",
+    "\n",
+    "clipped = mesh.clip(normal='y')\n",
+    "clipped_and_thresholded = clipped.threshold_percent(0.01)\n",
+    "clipped_and_thresholded.plot(volume=True)"
+   ]
+  },
   {
    "attachments": {},
    "cell_type": "markdown",
@@ -253,7 +397,7 @@
    "name": "python",
    "nbconvert_exporter": "python",
    "pygments_lexer": "ipython3",
-   "version": "3.10.12"
+   "version": "3.10.14"
   },
   "vscode": {
    "interpreter": {
diff --git a/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb
index de58d822..436de0f5 100644
--- a/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb
+++ b/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,
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/1_surface_dose_from_gamma_source.ipynb b/tasks/task_09_CSG_instantaneous_dose_tallies/1_surface_dose_from_gamma_source.ipynb
index f951da8e..a2a7e343 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/1_surface_dose_from_gamma_source.ipynb
+++ b/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",
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/2_surface_dose_from_gamma_source_study.ipynb b/tasks/task_09_CSG_instantaneous_dose_tallies/2_surface_dose_from_gamma_source_study.ipynb
index 57a937ce..cae8f30d 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/2_surface_dose_from_gamma_source_study.ipynb
+++ b/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,
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/3_cell_dose_from_neutron.py b/tasks/task_09_CSG_instantaneous_dose_tallies/3_cell_dose_from_neutron.py
index eccc3923..e4fc7e48 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/3_cell_dose_from_neutron.py
+++ b/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
 
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/4_cell_dose_from_photon.py b/tasks/task_09_CSG_instantaneous_dose_tallies/4_cell_dose_from_photon.py
index 260fbcfd..3834535d 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/4_cell_dose_from_photon.py
+++ b/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)
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/5_mesh_dose_from_neutrons.py b/tasks/task_09_CSG_instantaneous_dose_tallies/5_mesh_dose_from_neutrons.py
index c24cd5d3..1ade1c38 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/5_mesh_dose_from_neutrons.py
+++ b/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)
diff --git a/tasks/task_09_CSG_instantaneous_dose_tallies/compare_dose_simulation_with_back_of_envelope.py b/tasks/task_09_CSG_instantaneous_dose_tallies/compare_dose_simulation_with_back_of_envelope.py
index 1dbbbddf..48ff0a39 100644
--- a/tasks/task_09_CSG_instantaneous_dose_tallies/compare_dose_simulation_with_back_of_envelope.py
+++ b/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,
diff --git a/tasks/task_10_activation_transmutation_depletion/1_example_transmutation_isotope_build_up.ipynb b/tasks/task_10_activation_transmutation_depletion/1_example_transmutation_isotope_build_up.ipynb
index a323dcb0..0ad4a01b 100644
--- a/tasks/task_10_activation_transmutation_depletion/1_example_transmutation_isotope_build_up.ipynb
+++ b/tasks/task_10_activation_transmutation_depletion/1_example_transmutation_isotope_build_up.ipynb
@@ -30,14 +30,14 @@
     "\n",
     "import openmc\n",
     "import openmc.deplete\n",
-    "from pathlib import Path\n",
     "\n",
-    "# users might want to change these to use specific xml files to use particular decay data or transport cross sections\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'\n",
     "# the chain file was downloaded with\n",
     "# pip install openmc_data\n",
     "# download_endf_chain -r b8.0\n",
-    "# chain_file = '/nuclear_data/chain-endf-b8.0.xml'\n",
-    "# openmc.config['chain_file'] = chain_file\n"
+    "openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'\n"
    ]
   },
   {
diff --git a/tasks/task_10_activation_transmutation_depletion/2_example_tally_change_with_burnup.ipynb b/tasks/task_10_activation_transmutation_depletion/2_example_tally_change_with_burnup.ipynb
index 16f83e98..42cfefeb 100644
--- a/tasks/task_10_activation_transmutation_depletion/2_example_tally_change_with_burnup.ipynb
+++ b/tasks/task_10_activation_transmutation_depletion/2_example_tally_change_with_burnup.ipynb
@@ -9,21 +9,47 @@
     "This example simulates the variation of a tally responce as a function of time. This particular tally is the tritium breeding ratio and this tends to decrease over time as the lithium gets burnt up by neutron irradiation"
    ]
   },
+  {
+   "cell_type": "markdown",
+   "id": "9256b063",
+   "metadata": {},
+   "source": [
+    "\n",
+    "First import OpenMC and configure the nuclear data paths"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
-   "id": "9f65541a-65d9-423b-99f4-aa21f6e6b017",
+   "id": "b8b5587a",
    "metadata": {},
    "outputs": [],
    "source": [
     "# remove any old files\n",
-    "!rm settings.xm model.xml materials.xml geometry.xml settings.xml model.xml\n",
+    "!rm settings.xm model.xml materials.xml geometry.xml settings.xml\n",
     "\n",
     "import openmc\n",
     "import openmc.deplete\n",
     "import math\n",
     "\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'\n",
+    "# This chain file was downloaded using the download_endf_chain script that is included in the openmc_data package https://github.com/openmc-data-storage/openmc_data\\n\",\n",
+    "# this file tells openmc the decay paths between isotopes including probabilities of different routes and half lives\n",
+    "# To download this xml file you can run these commands\n",
+    "# pip install openmc_data\n",
+    "# download_endf_chain -d nuclear_data -r b8.0\n",
+    "openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "id": "9f65541a-65d9-423b-99f4-aa21f6e6b017",
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "mats = openmc.Materials()\n",
@@ -125,13 +151,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "# This chain file was downloaded using the download_endf_chain script that is included in the openmc_data package https://github.com/openmc-data-storage/openmc_data\\n\",\n",
-    "# this file tells openmc the decay paths between isotopes including probabilities of different routes and half lives\n",
-    "# To download this xml file you can run these commands\n",
-    "# pip install openmc_data\n",
-    "# download_endf_chain -d nuclear_data -r b8.0\n",
-    "# openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'\n",
-    "\n",
     "operator = openmc.deplete.CoupledOperator(\n",
     "    model=model,\n",
     "    normalization_mode=\"source-rate\",  # set for fixed source simulation, otherwise defaults to fission simulation\n",
diff --git a/tasks/task_10_activation_transmutation_depletion/3_full_pulse_schedule.ipynb b/tasks/task_10_activation_transmutation_depletion/3_full_pulse_schedule.ipynb
index b7b413d0..f7570e6c 100644
--- a/tasks/task_10_activation_transmutation_depletion/3_full_pulse_schedule.ipynb
+++ b/tasks/task_10_activation_transmutation_depletion/3_full_pulse_schedule.ipynb
@@ -31,22 +31,20 @@
     "# remove any old files\n",
     "!rm settings.xm model.xml materials.xml geometry.xml settings.xml\n",
     "\n",
+    "import matplotlib.pyplot as plt\n",
     "import openmc\n",
     "import openmc.deplete\n",
     "import math\n",
-    "import matplotlib.pyplot as plt"
-   ]
-  },
-  {
-   "cell_type": "code",
-   "execution_count": null,
-   "id": "f42c7e11-f514-4859-855f-873fb8f4026e",
-   "metadata": {},
-   "outputs": [],
-   "source": [
-    "# chain and cross section paths have been set on the docker image but you may want to change them\n",
-    "#openmc.config['chain_file']=path to chain file\n",
-    "#openmc.config['cross_sections']=path to cross_sections.xml"
+    "\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'\n",
+    "# This chain file was downloaded using the download_endf_chain script that is included in the openmc_data package https://github.com/openmc-data-storage/openmc_data\\n\",\n",
+    "# this file tells openmc the decay paths between isotopes including probabilities of different routes and half lives\n",
+    "# To download this xml file you can run these commands\n",
+    "# pip install openmc_data\n",
+    "# download_endf_chain -d nuclear_data -r b8.0\n",
+    "openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'"
    ]
   },
   {
diff --git a/tasks/task_13_stochastic_volume_calculation/1_find_volume_of_cell.ipynb b/tasks/task_13_stochastic_volume_calculation/1_find_volume_of_cell.ipynb
index 12e20d83..67564438 100644
--- a/tasks/task_13_stochastic_volume_calculation/1_find_volume_of_cell.ipynb
+++ b/tasks/task_13_stochastic_volume_calculation/1_find_volume_of_cell.ipynb
@@ -29,11 +29,11 @@
    ]
   },
   {
-   "attachments": {},
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "This first stage sets up the geometry and materials for the simulation."
+    "\n",
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -44,7 +44,25 @@
    "source": [
     "import openmc\n",
     "\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",
+   "metadata": {},
+   "source": [
+    "This first stage sets up the geometry and materials for the simulation."
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
     "# MATERIALS\n",
     "\n",
     "mat_iron = openmc.Material()\n",
diff --git a/tasks/task_14_variance_reduction/1_shielded_room_survival_biasing.ipynb b/tasks/task_14_variance_reduction/1_shielded_room_survival_biasing.ipynb
index f3c89010..d13f8804 100644
--- a/tasks/task_14_variance_reduction/1_shielded_room_survival_biasing.ipynb
+++ b/tasks/task_14_variance_reduction/1_shielded_room_survival_biasing.ipynb
@@ -16,11 +16,11 @@
    ]
   },
   {
-   "attachments": {},
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "First we import openmc and other packages needed for the example"
+    "\n",
+    "First import OpenMC and configure the nuclear data path"
    ]
   },
   {
@@ -31,7 +31,11 @@
    "source": [
     "import openmc\n",
     "from matplotlib import pyplot as plt\n",
-    "from matplotlib.colors import LogNorm  # used for plotting log scale graphs"
+    "from matplotlib.colors import LogNorm  # used for plotting log scale graphs\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'"
    ]
   },
   {
diff --git a/tasks/task_14_variance_reduction/2_shielded_room_single_ww.ipynb b/tasks/task_14_variance_reduction/2_shielded_room_single_ww.ipynb
index 390a357f..172d9c73 100644
--- a/tasks/task_14_variance_reduction/2_shielded_room_single_ww.ipynb
+++ b/tasks/task_14_variance_reduction/2_shielded_room_single_ww.ipynb
@@ -29,7 +29,7 @@
    "id": "f00eddb1-1e3a-4973-ba0c-d8feeb2a6704",
    "metadata": {},
    "source": [
-    "First we import ```openmc``` and other packages needed for the example"
+    "First we import ```openmc``` and other packages needed for the example and configure the nuclear data path"
    ]
   },
   {
@@ -40,13 +40,15 @@
    "outputs": [],
    "source": [
     "import time  # used to time the simulation\n",
-    "\n",
-    "import openmc\n",
+    "import numpy as np\n",
     "\n",
     "from matplotlib import pyplot as plt\n",
     "from matplotlib.colors import LogNorm  # used for plotting log scale graphs\n",
     "\n",
-    "import numpy as np"
+    "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'"
    ]
   },
   {
@@ -418,7 +420,7 @@
     "# we are reducing the number of particles so this simulation takes a similar amount of time\n",
     "# to the previous simulation that didn't use weight windows. So we can make a fair comparison\n",
     "model.settings.particles = 4900\n",
-    "model.settings.max_splits = 1_000  # controls the maximum partile splits over the entire lifetime of the particle\n",
+    "model.settings.max_history_splits = 1_000  # controls the maximum partile splits over the entire lifetime of the particle\n",
     "model.settings.weight_window_generators = wwg"
    ]
   },
diff --git a/tasks/task_14_variance_reduction/3_sphere_iterative_per_run_ww.ipynb b/tasks/task_14_variance_reduction/3_sphere_iterative_per_run_ww.ipynb
index 3f746188..4f7dd5d0 100644
--- a/tasks/task_14_variance_reduction/3_sphere_iterative_per_run_ww.ipynb
+++ b/tasks/task_14_variance_reduction/3_sphere_iterative_per_run_ww.ipynb
@@ -29,7 +29,7 @@
    "id": "f00eddb1-1e3a-4973-ba0c-d8feeb2a6704",
    "metadata": {},
    "source": [
-    "First we import ```openmc``` including ```openmc.lib``` and other packages needed for the example"
+    "First we import ```openmc``` including ```openmc.lib``` and other packages needed for the example. We also configure the nuclear data path."
    ]
   },
   {
@@ -39,11 +39,16 @@
    "metadata": {},
    "outputs": [],
    "source": [
+    "\n",
+    "from matplotlib import pyplot as plt\n",
+    "from matplotlib.colors import LogNorm  # used for plotting log scale graphs\n",
+    "\n",
     "import openmc\n",
     "import openmc.lib  # this example makes use of openmc lib to run the simulations\n",
     "\n",
-    "from matplotlib import pyplot as plt\n",
-    "from matplotlib.colors import LogNorm  # used for plotting log scale graphs"
+    "# 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'"
    ]
   },
   {
@@ -199,7 +204,7 @@
     "my_settings.source = source\n",
     "my_settings.particles = 120\n",
     "my_settings.batches = 10\n",
-    "my_settings.max_splits = 4000  # controls the total number of maximum splits a particle can do over the entire lifetime\n",
+    "my_settings.max_history_splits = 4000  # controls the total number of maximum splits a particle can do over the entire lifetime\n",
     "\n",
     "# no need to write the tallies.out file which saves space and time when large meshes are used\n",
     "my_settings.output = {'tallies': False}"
@@ -224,8 +229,8 @@
     "model = openmc.Model(my_geometry, my_materials, my_settings, my_tallies)\n",
     "\n",
     "# deletes old input and output files\n",
-    "!rm *.xml\n",
-    "!rm *.h5\n",
+    "!rm *.xml || true\n",
+    "!rm *.h5 || true\n",
     "\n",
     "model.export_to_xml()  # this is necessary as openmc.lib loads up the model.xml file"
    ]
diff --git a/tasks/task_14_variance_reduction/4_sphere_iterative_per_batch_ww.ipynb b/tasks/task_14_variance_reduction/4_sphere_iterative_per_batch_ww.ipynb
index 6889821c..2a6e5b3c 100644
--- a/tasks/task_14_variance_reduction/4_sphere_iterative_per_batch_ww.ipynb
+++ b/tasks/task_14_variance_reduction/4_sphere_iterative_per_batch_ww.ipynb
@@ -23,6 +23,14 @@
     "spectra tally."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "id": "61ad590c",
+   "metadata": {},
+   "source": [
+    "First we import the packages needed to run the example, including OpenMC. We also configure the nuclear data path."
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
@@ -30,6 +38,9 @@
    "metadata": {},
    "outputs": [],
    "source": [
+    "import numpy as np\n",
+    "import matplotlib.pyplot as plt\n",
+    "\n",
     "import openmc\n",
     "# Note this example makes use of OpenMC lib which provides python bindings to\n",
     "# the C/C++ methods in OpenMC and allows more direct control of the Monte Carlo\n",
@@ -37,8 +48,10 @@
     "# tally result each time.\n",
     "# Link to openmc.lib documentation https://docs.openmc.org/en/stable/pythonapi/capi.html\n",
     "import openmc.lib\n",
-    "import numpy as np\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'"
    ]
   },
   {
@@ -156,7 +169,7 @@
     "my_settings.source = source\n",
     "my_settings.particles = 50000\n",
     "my_settings.batches = 5\n",
-    "my_settings.max_splits = 100  # controls the total number of maximum splits a particle can do over the entire lifetime\n",
+    "my_settings.max_history_splits = 100  # controls the total number of maximum splits a particle can do over the entire lifetime\n",
     "\n",
     "# the mesh tallies produce large tallies.out files so this output setting avoids writing the tallies.out and saves time\n",
     "my_settings.output = {'tallies': False}"