Merge pull request #123 from fusion-energy/develop

general updates

.github/workflows/docker-publish.yml

@@ -58,6 +58,17 @@ jobs:
           rm -rf /tmp/.buildx-cache
           mv /tmp/.buildx-cache-new /tmp/.buildx-cache
 
+      - name: base build and push Docker image
+        uses: docker/build-push-action@v2
+        with:
+          push: true
+          target: base
+          tags: ghcr.io/fusion-energy/neutronics-workshop:base
+          build-args: |
+            include_avx=false
+            build_double_down=OFF
+            compile_cores=2
+
       - name: default build and push Docker image
         uses: docker/build-push-action@v2
         with:

tasks/task_01_cross_sections/1_isotope_xs_plot.ipynb

@@ -12,11 +12,11 @@
     "\n",
     "In this first part of the task we are plotting the <b>microscopic</b> cross-section \"<b>σ</b>\".\n",
     "\n",
-    "Microscopic cross section is the effective target area in m$^2$ presented by a single nucleus to an incident neutron beam.\n",
+    "Microscopic cross section is the effective target area in $\\mathrm{m}^2$ presented by a single nucleus to an incident neutron beam.\n",
     "\n",
     "This is used to characterize the probability of reaction between a neutron and an individual nucleus.\n",
     "\n",
-    "Microscopic is often stated in units of barns where 1 barn is equal to 10$^{−28}$ m$^2$ "
+    "Microscopic is often stated in units of barns where 1 barn is equal to $10^{−28}$ $\\mathrm{m}^2$ "
    ]
   },
   {

tasks/task_01_cross_sections/3_material_xs_plot.ipynb

@@ -16,9 +16,9 @@
     "\n",
     "Σ=σ.N\n",
     "\n",
-    "Where N is the number density of nuclei in units of nuclei / m$^3$.\n",
+    "Where N is the number density of nuclei in units of nuclei / $\\mathrm{m}^3$.\n",
     "\n",
-    "The units of Macroscopic cross section are therefore m$^{-1}$."
+    "The units of Macroscopic cross section are therefore $\\mathrm{m}^{-1}$."
    ]
   },
   {

tasks/task_07_CSG_cell_tally_spectra/1_example_neutron_spectra.ipynb

@@ -2,27 +2,24 @@
  "cells": [
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "# Part 1 - Plotting neutron energy spectra in a histogram form\n",
     "\n",
     "This example creates a simple sphere of water and tallies the neutrons in two different ways:\n",
     "\n",
     "- Neutron flux averaged across the cell.\n",
     "- Neutron current on the rear surface."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": 8,
-   "source": [
-    "from IPython.display import HTML\n",
-    "HTML('<iframe width=\"560\" height=\"315\" src=\"https://www.youtube.com/embed/qHqAuqMLYPA\" frameborder=\"0\" allow=\"accelerometer; autoplay; clipboard-write; encrypted-media; gyroscope; picture-in-picture\" allowfullscreen></iframe>')"
-   ],
+   "metadata": {},
    "outputs": [
     {
-     "output_type": "stream",
      "name": "stderr",
+     "output_type": "stream",
      "text": [
       "/home/jshimwell/miniconda3/envs/openmc-dagmc/lib/python3.8/site-packages/IPython/core/display.py:724: UserWarning:\n",
       "\n",
@@ -31,31 +28,36 @@
      ]
     },
     {
-     "output_type": "execute_result",
      "data": {
-      "text/plain": [
-       "<IPython.core.display.HTML object>"
-      ],
       "text/html": [
        "<iframe width=\"560\" height=\"315\" src=\"https://www.youtube.com/embed/qHqAuqMLYPA\" frameborder=\"0\" allow=\"accelerometer; autoplay; clipboard-write; encrypted-media; gyroscope; picture-in-picture\" allowfullscreen></iframe>"
+      ],
+      "text/plain": [
+       "<IPython.core.display.HTML object>"
       ]
      },
+     "execution_count": 8,
      "metadata": {},
-     "execution_count": 8
+     "output_type": "execute_result"
     }
    ],
-   "metadata": {}
+   "source": [
+    "from IPython.display import HTML\n",
+    "HTML('<iframe width=\"560\" height=\"315\" src=\"https://www.youtube.com/embed/qHqAuqMLYPA\" frameborder=\"0\" allow=\"accelerometer; autoplay; clipboard-write; encrypted-media; gyroscope; picture-in-picture\" allowfullscreen></iframe>')"
+   ]
   },
   {
    "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."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
     "import openmc\n",
     "\n",
@@ -107,27 +109,38 @@
     "source.angle = openmc.stats.Isotropic()\n",
     "source.energy = openmc.stats.Discrete([14e6], [1])\n",
     "sett.source = source"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "This section section adds a tally for the average neutron flux across a cell."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
+    "import numpy as np\n",
+    "\n",
     "#creates an empty tally object\n",
     "tallies = openmc.Tallies()\n",
     "\n",
     "# sets up filters for the tallies\n",
     "neutron_particle_filter = openmc.ParticleFilter(['neutron'])\n",
-    "energy_filter = openmc.EnergyFilter.from_group_structure('CCFE-709')\n",
+    "\n",
+    "# creates an array of 200 linear spaced energy bins from 0MeV to 15MeV\n",
+    "# our source is 14MeV so this should capture all the neutron energies in the simulation\n",
+    "energy_filter = openmc.EnergyFilter(np.linspace(0,15e6,200))\n",
+    "\n",
+    "# alternatively you could use one of the inbuilt energy group structures\n",
+    "# these are not linearly spaced as they have extra bins in key energy ranges\n",
+    "# A full list of energy structures is available here\n",
+    "# https://github.com/openmc-dev/openmc/blob/6254be37582e09acff038f5656332b89e53e4eae/openmc/mgxs/__init__.py#L50-L420\n",
+    "# energy_filter = openmc.EnergyFilter.from_group_structure('CCFE-709')\n",
     "\n",
     "\n",
     "# setup the filters for the cell tally\n",
@@ -138,24 +151,23 @@
     "cell_spectra_tally.scores = ['flux']\n",
     "cell_spectra_tally.filters = [cell_filter, neutron_particle_filter, energy_filter]\n",
     "tallies.append(cell_spectra_tally)"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "This section adds two surface current tallies - one on the inner sphere surface and one on the outer sphere surface."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
     "# sets up filters for the tallies\n",
-    "neutron_particle_filter = openmc.ParticleFilter(['neutron'])\n",
-    "energy_filter = openmc.EnergyFilter.from_group_structure('CCFE-709')\n",
+    "# neutron_particle_filter and energy_filter are used from the previous cell\n",
     "\n",
     "# setup the filters for the surface tally\n",
     "front_surface_filter = openmc.SurfaceFilter(vessel_inner_surface)\n",
@@ -171,49 +183,49 @@
     "back_surface_spectra_tally.scores = ['current']\n",
     "back_surface_spectra_tally.filters = [back_surface_filter, neutron_particle_filter, energy_filter]\n",
     "tallies.append(back_surface_spectra_tally)"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "This section runs the simulation."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
     "# combine all the required parts to make a model\n",
     "model = openmc.model.Model(geom, mats, sett, tallies)\n",
     "\n",
     "# remove old files and runs OpenMC\n",
     "!rm *.h5\n",
     "results_filename = model.run()"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "This section extracts the cell tally data from the results file and plots neutron flux across the cell."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
     "# open the results file\n",
     "results = openmc.StatePoint(results_filename)\n",
     "\n",
     "#extracts the tally values from the simulation results\n",
     "cell_tally = results.get_tally(name='cell_spectra_tally')\n",
     "\n",
-    "# imports a convenient plotting package\n",
+    "# imports a convenient plotting package, but you can extract the raw results from cell_tally directly\n",
     "from spectrum_plotter import plot_spectrum_from_tally\n",
     "\n",
     "plot_spectrum_from_tally(\n",
@@ -222,20 +234,20 @@
     "    y_label = \"flux [neutrons centimeter / source neutron]\",\n",
     "    plotting_package=\"plotly\",  # matplotlib is another option, but plotly option provides a dropdown menu for changing axis scale\n",
     ")"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "This section extracts the surface tally data from the results file and plots neutron current through the inner and outer surfaces."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
+   "metadata": {},
+   "outputs": [],
    "source": [
     "# open the results file\n",
     "results = openmc.StatePoint(results_filename)\n",
@@ -244,11 +256,6 @@
     "back_surface_tally = results.get_tally(name='back_surface_spectra_tally')\n",
     "front_surface_tally = results.get_tally(name='front_surface_spectra_tally')\n",
     "\n",
-    "# imports a convenient plotting package\n",
-    "from spectrum_plotter import plot_spectrum_from_tally\n",
-    "\n",
-    "# import openmc_post_processor as opp\n",
-    "# opp.get_tally_units(front_surface_tally)\n",
     "\n",
     "plot_spectrum_from_tally(\n",
     "    spectrum={'back_surface_tally': back_surface_tally, 'front_surface_tally':front_surface_tally},\n",
@@ -257,12 +264,11 @@
     "    plotting_package=\"plotly\",  # matplotlib is another option, but plotly option provides a dropdown menu for changing axis scale\n",
     "    required_units = 'neutrons / source_particle'\n",
     ")"
-   ],
-   "outputs": [],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "#### Additional notes\n",
     "\n",
@@ -271,25 +277,24 @@
     "- When comparing these results with other codes it might be necessary to scale them:\n",
     "    - The OpenMC surface current needs to be divided by the surface area when comparing with MCNP.\n",
     "    - The OpenMC cell flux needs to be divided by the cell volume when comparing with MCNP.\n"
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "markdown",
+   "metadata": {},
    "source": [
     "**Learning Outcomes for Part 1:**\n",
     "\n",
     "- Neutron spectra can be tallied in OpenMC using flux or current tallies, and energy bins.\n",
     "- Neutron spectra tend to be different on different surfaces and cells in the geometry."
-   ],
-   "metadata": {}
+   ]
   },
   {
    "cell_type": "code",
    "execution_count": null,
-   "source": [],
+   "metadata": {},
    "outputs": [],
-   "metadata": {}
+   "source": []
   }
  ],
  "metadata": {
@@ -316,4 +321,4 @@
  },
  "nbformat": 4,
  "nbformat_minor": 4
-}
+}