From 5e313cf5f1d601074ad95c17ae589bf564972adb Mon Sep 17 00:00:00 2001 From: Adam Nelson Date: Sun, 26 Feb 2017 05:33:32 -0500 Subject: [PATCH] Revisions per @wbinventors comments --- docs/source/pythonapi/examples/mgxs-part-i.ipynb | 4 +++- 1 file changed, 3 insertions(+), 1 deletion(-) diff --git a/docs/source/pythonapi/examples/mgxs-part-i.ipynb b/docs/source/pythonapi/examples/mgxs-part-i.ipynb index 2b3d5a77c..17d64ee2d 100644 --- a/docs/source/pythonapi/examples/mgxs-part-i.ipynb +++ b/docs/source/pythonapi/examples/mgxs-part-i.ipynb @@ -380,7 +380,9 @@ "* `InverseVelocity`\n", "* `PromptNuFissionXS`\n", "\n", - "Of course, we know that the transport (`TransportXS`), fission (`FissionXS`), scattering (`ScatterXS`), and scattering-matrix (`ScatterMatrixXS`) cross sections can potentially incorporate neutron multiplication ($\\nu$). For these types, the multpiplication can be accomodated by setting the `nu` parameter to `True` as shown below.\n", + "Of course, we are aware that the fission cross section (`FissionXS`) can sometimes be paired with the fission neutron multiplication to become $\\nu\\sigma_f$. This can be accomodated in to the `FissionXS` class by setting the `nu` parameter to `True` as shown below.\n", + "\n", + "Additionally, scattering reactions (like (n,2n)) can also be defined to take in to account the neutron multiplication to become $\\nu\\sigma_s$. This can be accomodated in the the transport (`TransportXS`), scattering (`ScatterXS`), and scattering-matrix (`ScatterMatrixXS`) cross sections types by setting the `nu` parameter to `True` as shown below.\n", "\n", "These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure." ]