added nuc data path

fusion-energy · Oct 17, 2024 · 5a740a5 · 5a740a5
1 parent 2df1f0e
commit 5a740a5
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Showing 5 changed files with 88 additions and 9 deletions.
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
@@ -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",

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
@@ -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",

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
@@ -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",

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
@@ -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",

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
@@ -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'"
    ]
   },
   {