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Updated MGXS Notebooks per comments from @paulromano

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wbinventor@gmail.com 2015-12-02 09:18:19 -05:00
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docs/source/pythonapi/examples/mgxs-part-i.ipynb
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"cell_type": "markdown",
"metadata": {},
"source": [
"Many Monte Carlo-based neutron particle transport codes, including OpenMC, use continuous energy nuclear cross section data. However, most deterministic neutron transport codes use *multi-group cross sections* defined over discretized energy bins or *energy groups*. An example of U-235's fission continuous energy cross section along with a 16-group cross section computed for a light water reactor spectrum is displayed below."
"Many Monte Carlo particle transport codes, including OpenMC, use continuous-energy nuclear cross section data. However, most deterministic neutron transport codes use *multi-group cross sections* defined over discretized energy bins or *energy groups*. An example of U-235's continuous-energy fission cross section along with a 16-group cross section computed for a light water reactor spectrum is displayed below."
]
},
{
@ -79,7 +79,7 @@
"### Spatial and Energy Discretization\n",
"The energy domain for critical systems such as thermal reactors spans more than 10 orders of magnitude of neutron energies from 10$^{-5}$ - 10$^7$ eV. The multi-group approximation discretization divides this energy range into one or more energy groups. In particular, for $G$ total groups, we denote an energy group index $g$ such that $g \\in \\{1, 2, ..., G\\}$. The energy group indices are defined such that the smaller group the higher the energy, and vice versa. The integration over neutron energies across a discrete energy group is commonly referred to as **energy condensation**.\n",
"\n",
"Multi-group cross sections are computed for discretized spatial zones in the geometry of interest. The spatial zones may be defined on a structured and regular fuel assembly or pin cell mesh, or an unstructured mesh such as the constructive solid geometry used by OpenMC. For a geometry with $K$ distinct spatial zones, we designate each spatial zone an index $k$ such that $k \\in \\{1, 2, ..., K\\}$. The volume of each spatial zone is denoted by $V_{k}$. The integration over discrete spatial zones is commonly referred to as **spatial homogenization**."
"Multi-group cross sections are computed for discretized spatial zones in the geometry of interest. The spatial zones may be defined on a structured and regular fuel assembly or pin cell mesh, an arbitrary unstructured mesh or the constructive solid geometry used by OpenMC. For a geometry with $K$ distinct spatial zones, we designate each spatial zone an index $k$ such that $k \\in \\{1, 2, ..., K\\}$. The volume of each spatial zone is denoted by $V_{k}$. The integration over discrete spatial zones is commonly referred to as **spatial homogenization**."
]
},
{
@ -518,7 +518,7 @@
" License: http://mit-crpg.github.io/openmc/license.html\n",
" Version: 0.7.0\n",
" Git SHA1: c4b14a5ef87f004528d35cbf33fef3ed15a386ca\n",
" Date/Time: 2015-11-30 21:26:04\n",
" Date/Time: 2015-12-02 09:11:05\n",
" MPI Processes: 1\n",
"\n",
" ===========================================================================\n",
@ -605,19 +605,19 @@
" =======================> TIMING STATISTICS <=======================\n",
"\n",
" Total time for initialization = 4.1700E-01 seconds\n",
" Reading cross sections = 9.7000E-02 seconds\n",
" Total time in simulation = 1.4656E+01 seconds\n",
" Time in transport only = 1.4643E+01 seconds\n",
" Time in inactive batches = 1.7940E+00 seconds\n",
" Time in active batches = 1.2862E+01 seconds\n",
" Reading cross sections = 8.9000E-02 seconds\n",
" Total time in simulation = 1.4728E+01 seconds\n",
" Time in transport only = 1.4712E+01 seconds\n",
" Time in inactive batches = 1.7890E+00 seconds\n",
" Time in active batches = 1.2939E+01 seconds\n",
" Time synchronizing fission bank = 5.0000E-03 seconds\n",
" Sampling source sites = 4.0000E-03 seconds\n",
" SEND/RECV source sites = 1.0000E-03 seconds\n",
" Time accumulating tallies = 0.0000E+00 seconds\n",
" Total time for finalization = 0.0000E+00 seconds\n",
" Total time elapsed = 1.5082E+01 seconds\n",
" Calculation Rate (inactive) = 13935.3 neutrons/second\n",
" Calculation Rate (active) = 7774.84 neutrons/second\n",
" Sampling source sites = 3.0000E-03 seconds\n",
" SEND/RECV source sites = 2.0000E-03 seconds\n",
" Time accumulating tallies = 1.0000E-03 seconds\n",
" Total time for finalization = 1.0000E-03 seconds\n",
" Total time elapsed = 1.5155E+01 seconds\n",
" Calculation Rate (inactive) = 13974.3 neutrons/second\n",
" Calculation Rate (active) = 7728.57 neutrons/second\n",
"\n",
" ============================> RESULTS <============================\n",
"\n",
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"collapsed": false
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"name": "stderr",
"output_type": "stream",
"text": [
"/usr/local/lib/python2.7/dist-packages/openmc-0.7.0-py2.7.egg/openmc/mgxs/mgxs.py:1254: FutureWarning: elementwise comparison failed; returning scalar instead, but in the future will perform elementwise comparison\n"
]
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docs/source/pythonapi/examples/mgxs-part-iii.ipynb
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"cell_type": "markdown",
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"source": [
"Now, we must specify to the `Library` which types of cross sections to compute. In particular, the following are the multi-group cross section `MGXS` subclasses are mapped to string codes accepted by the `Library` class:\n",
"Now, we must specify to the `Library` which types of cross sections to compute. In particular, the following are the multi-group cross section `MGXS` subclasses that are mapped to string codes accepted by the `Library` class:\n",
"\n",
"* `TotalXS` (`\"total\"`)\n",
"* `TransportXS` (`\"transport\"`)\n",
@ -580,7 +580,7 @@
"source": [
"Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports `\"material,\"` `\"cell,\"` and `\"universe\"` domain types. We will use a `\"cell\"` domain type here to compute cross sections in each of the cells in the fuel assembly geometry.\n",
"\n",
"**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property. In our case, we wish to compute multi-group cross sectoins in each and every cell since they will be needed in our downstream OpenMOC calculation on the identical combinatorial geometry mesh."
"**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property. In our case, we wish to compute multi-group cross sections in each and every cell since they will be needed in our downstream OpenMOC calculation on the identical combinatorial geometry mesh."
]
},
{