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Revisions per @wbinventors comments

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Adam Nelson 2017-02-26 05:33:32 -05:00
parent c40a1ca2fd
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4
docs/source/pythonapi/examples/mgxs-part-i.ipynb
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@ -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."
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