From 31c1d39078b8320028128a33f5b83dc01e7c3379 Mon Sep 17 00:00:00 2001 From: Will Boyd Date: Wed, 28 Oct 2015 08:22:16 -0400 Subject: [PATCH] Updated ipython notebook for mgxs per comments from @paulromano --- .../examples/multi-group-cross-sections.ipynb | 10 +++++----- 1 file changed, 5 insertions(+), 5 deletions(-) diff --git a/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb b/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb index cfbc68d39..a6dda13d6 100644 --- a/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb +++ b/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb @@ -288,7 +288,7 @@ "cell_type": "markdown", "metadata": {}, "source": [ - "We can now use the fine and coarse `EnergyGroups` objects, along with our previously created materials and geometry, to instantiate some `MGXS` objects from the `openmc.mgxs` module. In particular, the following are subclasses of generic and abstract `MGXS` class:\n", + "We can now use the fine and coarse `EnergyGroups` objects, along with our previously created materials and geometry, to instantiate some `MGXS` objects from the `openmc.mgxs` module. In particular, the following are subclasses of the generic and abstract `MGXS` class:\n", "\n", "* `TotalXS`\n", "* `TransportXS`\n", @@ -666,7 +666,7 @@ "cell_type": "markdown", "metadata": {}, "source": [ - "The multi-group cross section objects can now use OpenMC's [tally arithmetic](http://mit-crpg.github.io/openmc/pythonapi/examples/pandas-dataframes.html) to compute cross sections from the tally data." + "The multi-group cross section objects can now use OpenMC's [tally arithmetic](http://mit-crpg.github.io/openmc/pythonapi/examples/tally-arithmetic.html) to compute cross sections from the tally data." ] }, { @@ -1443,7 +1443,7 @@ "\n", "# Add the tally trigger to each of the multi-group cross section tallies\n", "for cell in openmc_cells:\n", - " for mgxs_type in xs_library[cell.id].keys():\n", + " for mgxs_type in xs_library[cell.id]:\n", " xs_library[cell.id][mgxs_type].tally_trigger = tally_trigger\n", " \n", "# Set the trigger to active in the \"settings.xml\" file\n", @@ -1473,7 +1473,7 @@ "\n", "# Iterate over all cells and cross section types\n", "for cell in openmc_cells:\n", - " for rxn_type in xs_library[cell.id].keys():\n", + " for rxn_type in xs_library[cell.id]:\n", "\n", " # Set the cross sections domain type to the cell\n", " xs_library[cell.id][rxn_type].domain = cell\n", @@ -1725,7 +1725,7 @@ "source": [ "# Iterate over all cells and cross section types\n", "for cell in openmc_cells:\n", - " for rxn_type in xs_library[cell.id].keys():\n", + " for rxn_type in xs_library[cell.id]:\n", " xs_library[cell.id][rxn_type].load_from_statepoint(sp)\n", " xs_library[cell.id][rxn_type].compute_xs()" ]