mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-27 05:35:49 -04:00
Updated docstrings for MultiGroupXS subclasses in Python API
This commit is contained in:
Will Boyd
parent
3baaacda99
commit
40f893e6d0
3 changed files with 813 additions and 438 deletions
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@ -369,7 +369,7 @@
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"outputs": [
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"text/plain": [
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"<IPython.core.display.Image object>"
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]
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@ -580,7 +580,7 @@
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" License: http://mit-crpg.github.io/openmc/license.html\n",
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" Version: 0.7.0\n",
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" Git SHA1: e0c2aace2e73367536fa03e153b67a2d038cd2b3\n",
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" Date/Time: 2015-10-03 01:14:27\n",
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" Date/Time: 2015-10-03 02:50:47\n",
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" MPI Processes: 1\n",
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"\n",
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" ===========================================================================\n",
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@ -636,20 +636,20 @@
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"\n",
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" =======================> TIMING STATISTICS <=======================\n",
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"\n",
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" Total time for initialization = 6.7400E-01 seconds\n",
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" Reading cross sections = 1.5200E-01 seconds\n",
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" Total time in simulation = 2.4330E+01 seconds\n",
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" Time in transport only = 2.4308E+01 seconds\n",
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" Time in inactive batches = 2.4220E+00 seconds\n",
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" Time in active batches = 2.1908E+01 seconds\n",
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" Total time for initialization = 1.1480E+00 seconds\n",
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" Reading cross sections = 2.9100E-01 seconds\n",
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" Total time in simulation = 2.7345E+01 seconds\n",
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" Time in transport only = 2.7286E+01 seconds\n",
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" Time in inactive batches = 5.8310E+00 seconds\n",
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" Time in active batches = 2.1514E+01 seconds\n",
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" Time synchronizing fission bank = 1.0000E-03 seconds\n",
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" Sampling source sites = 1.0000E-03 seconds\n",
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" SEND/RECV source sites = 0.0000E+00 seconds\n",
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" Time accumulating tallies = 0.0000E+00 seconds\n",
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" Total time for finalization = 1.0000E-03 seconds\n",
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" Total time elapsed = 2.5018E+01 seconds\n",
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" Calculation Rate (inactive) = 5161.02 neutrons/second\n",
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" Calculation Rate (active) = 1711.70 neutrons/second\n",
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" Total time for finalization = 2.0000E-03 seconds\n",
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" Total time elapsed = 2.8526E+01 seconds\n",
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" Calculation Rate (inactive) = 2143.71 neutrons/second\n",
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" Calculation Rate (active) = 1743.05 neutrons/second\n",
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"\n",
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" ============================> RESULTS <============================\n",
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"\n",
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@ -721,20 +721,7 @@
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"collapsed": false,
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"scrolled": true
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},
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"outputs": [
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{
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"ename": "KeyError",
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"evalue": "10003",
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"output_type": "error",
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"traceback": [
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"\u001b[1;31m---------------------------------------------------------------------------\u001b[0m",
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||||
"\u001b[1;31mKeyError\u001b[0m Traceback (most recent call last)",
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||||
"\u001b[1;32m<ipython-input-25-3cd87b8121ef>\u001b[0m in \u001b[0;36m<module>\u001b[1;34m()\u001b[0m\n\u001b[0;32m 1\u001b[0m \u001b[1;31m# Load the summary file and link with statepoint\u001b[0m\u001b[1;33m\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[0;32m 2\u001b[0m \u001b[0msu\u001b[0m \u001b[1;33m=\u001b[0m \u001b[0mSummary\u001b[0m\u001b[1;33m(\u001b[0m\u001b[1;34m'summary.h5'\u001b[0m\u001b[1;33m)\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[1;32m----> 3\u001b[1;33m \u001b[0msp\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mlink_with_summary\u001b[0m\u001b[1;33m(\u001b[0m\u001b[0msu\u001b[0m\u001b[1;33m)\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[0m",
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||||
"\u001b[1;32m/usr/local/lib/python2.7/dist-packages/openmc-0.7.0-py2.7.egg/openmc/statepoint.pyc\u001b[0m in \u001b[0;36mlink_with_summary\u001b[1;34m(self, summary)\u001b[0m\n\u001b[0;32m 610\u001b[0m \u001b[1;32mfor\u001b[0m \u001b[0mtally_id\u001b[0m\u001b[1;33m,\u001b[0m \u001b[0mtally\u001b[0m \u001b[1;32min\u001b[0m \u001b[0mself\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mtallies\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mitems\u001b[0m\u001b[1;33m(\u001b[0m\u001b[1;33m)\u001b[0m\u001b[1;33m:\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[0;32m 611\u001b[0m \u001b[1;31m# Get the Tally name from the summary file\u001b[0m\u001b[1;33m\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[1;32m--> 612\u001b[1;33m \u001b[0mtally\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mname\u001b[0m \u001b[1;33m=\u001b[0m \u001b[0msummary\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mtallies\u001b[0m\u001b[1;33m[\u001b[0m\u001b[0mtally_id\u001b[0m\u001b[1;33m]\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mname\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[0m\u001b[0;32m 613\u001b[0m \u001b[0mtally\u001b[0m\u001b[1;33m.\u001b[0m\u001b[0mwith_summary\u001b[0m \u001b[1;33m=\u001b[0m \u001b[0mTrue\u001b[0m\u001b[1;33m\u001b[0m\u001b[0m\n\u001b[0;32m 614\u001b[0m \u001b[1;33m\u001b[0m\u001b[0m\n",
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||||
"\u001b[1;31mKeyError\u001b[0m: 10003"
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||||
]
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||||
}
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||||
],
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"outputs": [],
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"source": [
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"# Load the summary file and link with statepoint\n",
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"su = Summary('summary.h5')\n",
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@ -752,11 +739,47 @@
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},
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{
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||||
"cell_type": "code",
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"execution_count": null,
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||||
"execution_count": 26,
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||||
"metadata": {
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||||
"collapsed": false
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||||
},
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||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
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||||
" <tr style=\"text-align: right;\">\n",
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" <th></th>\n",
|
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" <th>nuclide</th>\n",
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" <th>score</th>\n",
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" <th>mean</th>\n",
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" <th>std. dev.</th>\n",
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||||
" </tr>\n",
|
||||
" </thead>\n",
|
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" <tbody>\n",
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" <tr>\n",
|
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" <th>0</th>\n",
|
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" <td>total</td>\n",
|
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" <td>(nu-fission / absorption)</td>\n",
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" <td>1.046353</td>\n",
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" <td>0.00935</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
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||||
"</div>"
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||||
],
|
||||
"text/plain": [
|
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" nuclide score mean std. dev.\n",
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||||
"0 total (nu-fission / absorption) 1.046353 0.00935"
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]
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||||
},
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"execution_count": 26,
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"metadata": {},
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||||
"output_type": "execute_result"
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}
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],
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"source": [
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"# Compute k-infinity using tally arithmetic\n",
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"fiss_rate = sp.get_tally(name='fiss. rate')\n",
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@ -776,11 +799,49 @@
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},
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{
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"cell_type": "code",
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"execution_count": null,
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"execution_count": 27,
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"metadata": {
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"collapsed": false
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},
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||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
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" <thead>\n",
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" <tr style=\"text-align: right;\">\n",
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" <th></th>\n",
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" <th>energy [MeV]</th>\n",
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" <th>nuclide</th>\n",
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" <th>score</th>\n",
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" <th>mean</th>\n",
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" <th>std. dev.</th>\n",
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" </tr>\n",
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" </thead>\n",
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" <tbody>\n",
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" <tr>\n",
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" <th>0</th>\n",
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" <td>(0.0e+00 - 6.2e-01)</td>\n",
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" <td>total</td>\n",
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" <td>absorption</td>\n",
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" <td>0.95873</td>\n",
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" <td>0.00774</td>\n",
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" </tr>\n",
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||||
" </tbody>\n",
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||||
"</table>\n",
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||||
"</div>"
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||||
],
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||||
"text/plain": [
|
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" energy [MeV] nuclide score mean std. dev.\n",
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"0 (0.0e+00 - 6.2e-01) total absorption 0.95873 0.00774"
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]
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},
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||||
"execution_count": 27,
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||||
"metadata": {},
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||||
"output_type": "execute_result"
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}
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],
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"source": [
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"# Compute resonance escape probability using tally arithmetic\n",
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"therm_abs_rate = sp.get_tally(name='therm. abs. rate')\n",
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@ -798,11 +859,47 @@
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},
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{
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"cell_type": "code",
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"execution_count": null,
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"execution_count": 28,
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"metadata": {
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"collapsed": false
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},
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||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
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|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
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" <thead>\n",
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" <tr style=\"text-align: right;\">\n",
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" <th></th>\n",
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" <th>nuclide</th>\n",
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" <th>score</th>\n",
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" <th>mean</th>\n",
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" <th>std. dev.</th>\n",
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||||
" </tr>\n",
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||||
" </thead>\n",
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" <tbody>\n",
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" <tr>\n",
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" <th>0</th>\n",
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" <td>total</td>\n",
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" <td>nu-fission</td>\n",
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" <td>1.091622</td>\n",
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" <td>0.011163</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
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||||
],
|
||||
"text/plain": [
|
||||
" nuclide score mean std. dev.\n",
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||||
"0 total nu-fission 1.091622 0.011163"
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||||
]
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||||
},
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||||
"execution_count": 28,
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||||
"metadata": {},
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||||
"output_type": "execute_result"
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||||
}
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||||
],
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||||
"source": [
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"# Compute fast fission factor factor using tally arithmetic\n",
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"therm_fiss_rate = sp.get_tally(name='therm. fiss. rate')\n",
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|
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@ -821,11 +918,51 @@
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},
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{
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"cell_type": "code",
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"execution_count": null,
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||||
"execution_count": 29,
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||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
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||||
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||||
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|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
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" <th>energy [MeV]</th>\n",
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" <th>cell</th>\n",
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" <th>nuclide</th>\n",
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" <th>score</th>\n",
|
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" <th>mean</th>\n",
|
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" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
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" <tr>\n",
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" <th>0</th>\n",
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" <td>(0.0e+00 - 6.2e-01)</td>\n",
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" <td>10000</td>\n",
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" <td>total</td>\n",
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" <td>absorption</td>\n",
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" <td>0.802012</td>\n",
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||||
" <td>0.006609</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" energy [MeV] cell nuclide score mean std. dev.\n",
|
||||
"0 (0.0e+00 - 6.2e-01) 10000 total absorption 0.802012 0.006609"
|
||||
]
|
||||
},
|
||||
"execution_count": 29,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# Compute thermal flux utilization factor using tally arithmetic\n",
|
||||
"fuel_therm_abs_rate = sp.get_tally(name='fuel therm. abs. rate')\n",
|
||||
|
|
@ -842,11 +979,49 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 30,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
||||
" <th>energy [MeV]</th>\n",
|
||||
" <th>nuclide</th>\n",
|
||||
" <th>score</th>\n",
|
||||
" <th>mean</th>\n",
|
||||
" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
||||
" <tr>\n",
|
||||
" <th>0</th>\n",
|
||||
" <td>(0.0e+00 - 6.2e-01)</td>\n",
|
||||
" <td>total</td>\n",
|
||||
" <td>(nu-fission / absorption)</td>\n",
|
||||
" <td>1.246604</td>\n",
|
||||
" <td>0.011825</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" energy [MeV] nuclide score mean std. dev.\n",
|
||||
"0 (0.0e+00 - 6.2e-01) total (nu-fission / absorption) 1.246604 0.011825"
|
||||
]
|
||||
},
|
||||
"execution_count": 30,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# Compute neutrons produced per absorption (eta) using tally arithmetic\n",
|
||||
"eta = therm_fiss_rate / fuel_therm_abs_rate\n",
|
||||
|
|
@ -862,11 +1037,52 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 31,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
||||
" <th>energy [MeV]</th>\n",
|
||||
" <th>nuclide</th>\n",
|
||||
" <th>score</th>\n",
|
||||
" <th>mean</th>\n",
|
||||
" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
||||
" <tr>\n",
|
||||
" <th>0</th>\n",
|
||||
" <td>(0.0e+00 - 6.2e-01)</td>\n",
|
||||
" <td>total</td>\n",
|
||||
" <td>(((absorption * nu-fission) * absorption) * (n...</td>\n",
|
||||
" <td>1.046353</td>\n",
|
||||
" <td>0.01894</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" energy [MeV] nuclide \\\n",
|
||||
"0 (0.0e+00 - 6.2e-01) total \n",
|
||||
"\n",
|
||||
" score mean std. dev. \n",
|
||||
"0 (((absorption * nu-fission) * absorption) * (n... 1.046353 0.01894 "
|
||||
]
|
||||
},
|
||||
"execution_count": 31,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"keff = res_esc * fast_fiss * therm_util * eta\n",
|
||||
"keff.get_pandas_dataframe()"
|
||||
|
|
@ -883,7 +1099,7 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 32,
|
||||
"metadata": {
|
||||
"collapsed": false,
|
||||
"scrolled": true
|
||||
|
|
@ -899,11 +1115,131 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 33,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
||||
" <th>cell</th>\n",
|
||||
" <th>energy [MeV]</th>\n",
|
||||
" <th>nuclide</th>\n",
|
||||
" <th>score</th>\n",
|
||||
" <th>mean</th>\n",
|
||||
" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
||||
" <tr>\n",
|
||||
" <th>0</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>(U-238 / total)</td>\n",
|
||||
" <td>(nu-fission / flux)</td>\n",
|
||||
" <td>6.641746e-07</td>\n",
|
||||
" <td>6.859257e-09</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>1</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>(U-238 / total)</td>\n",
|
||||
" <td>(scatter / flux)</td>\n",
|
||||
" <td>2.099861e-01</td>\n",
|
||||
" <td>1.966887e-03</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>2</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>(U-235 / total)</td>\n",
|
||||
" <td>(nu-fission / flux)</td>\n",
|
||||
" <td>3.556665e-01</td>\n",
|
||||
" <td>3.717881e-03</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>3</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>(U-235 / total)</td>\n",
|
||||
" <td>(scatter / flux)</td>\n",
|
||||
" <td>5.554650e-03</td>\n",
|
||||
" <td>5.218094e-05</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>4</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>(U-238 / total)</td>\n",
|
||||
" <td>(nu-fission / flux)</td>\n",
|
||||
" <td>7.165057e-03</td>\n",
|
||||
" <td>5.625590e-05</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>5</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>(U-238 / total)</td>\n",
|
||||
" <td>(scatter / flux)</td>\n",
|
||||
" <td>2.276535e-01</td>\n",
|
||||
" <td>8.544314e-04</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>6</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>(U-235 / total)</td>\n",
|
||||
" <td>(nu-fission / flux)</td>\n",
|
||||
" <td>8.089493e-03</td>\n",
|
||||
" <td>5.080374e-05</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>7</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>(U-235 / total)</td>\n",
|
||||
" <td>(scatter / flux)</td>\n",
|
||||
" <td>3.370111e-03</td>\n",
|
||||
" <td>1.361116e-05</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" cell energy [MeV] nuclide score \\\n",
|
||||
"0 10000 (0.0e+00 - 6.3e-07) (U-238 / total) (nu-fission / flux) \n",
|
||||
"1 10000 (0.0e+00 - 6.3e-07) (U-238 / total) (scatter / flux) \n",
|
||||
"2 10000 (0.0e+00 - 6.3e-07) (U-235 / total) (nu-fission / flux) \n",
|
||||
"3 10000 (0.0e+00 - 6.3e-07) (U-235 / total) (scatter / flux) \n",
|
||||
"4 10000 (6.3e-07 - 2.0e+01) (U-238 / total) (nu-fission / flux) \n",
|
||||
"5 10000 (6.3e-07 - 2.0e+01) (U-238 / total) (scatter / flux) \n",
|
||||
"6 10000 (6.3e-07 - 2.0e+01) (U-235 / total) (nu-fission / flux) \n",
|
||||
"7 10000 (6.3e-07 - 2.0e+01) (U-235 / total) (scatter / flux) \n",
|
||||
"\n",
|
||||
" mean std. dev. \n",
|
||||
"0 6.641746e-07 6.859257e-09 \n",
|
||||
"1 2.099861e-01 1.966887e-03 \n",
|
||||
"2 3.556665e-01 3.717881e-03 \n",
|
||||
"3 5.554650e-03 5.218094e-05 \n",
|
||||
"4 7.165057e-03 5.625590e-05 \n",
|
||||
"5 2.276535e-01 8.544314e-04 \n",
|
||||
"6 8.089493e-03 5.080374e-05 \n",
|
||||
"7 3.370111e-03 1.361116e-05 "
|
||||
]
|
||||
},
|
||||
"execution_count": 33,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"fuel_xs = fuel_rxn_rates / flux\n",
|
||||
"fuel_xs.get_pandas_dataframe()"
|
||||
|
|
@ -918,11 +1254,23 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 34,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"name": "stdout",
|
||||
"output_type": "stream",
|
||||
"text": [
|
||||
"[[[ 6.64174599e-07]\n",
|
||||
" [ 3.55666541e-01]]\n",
|
||||
"\n",
|
||||
" [[ 7.16505734e-03]\n",
|
||||
" [ 8.08949336e-03]]]\n"
|
||||
]
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# Show how to use Tally.get_values(...) with a CrossScore\n",
|
||||
"nu_fiss_xs = fuel_xs.get_values(scores=['(nu-fission / flux)'])\n",
|
||||
|
|
@ -938,11 +1286,21 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 35,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"name": "stdout",
|
||||
"output_type": "stream",
|
||||
"text": [
|
||||
"[[[ 0.00555465]]\n",
|
||||
"\n",
|
||||
" [[ 0.00337011]]]\n"
|
||||
]
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# Show how to use Tally.get_values(...) with a CrossScore and CrossNuclide\n",
|
||||
"u235_scatter_xs = fuel_xs.get_values(nuclides=['(U-235 / total)'], \n",
|
||||
|
|
@ -952,11 +1310,20 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 36,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"name": "stdout",
|
||||
"output_type": "stream",
|
||||
"text": [
|
||||
"[[[ 0.22765348]\n",
|
||||
" [ 0.00337011]]]\n"
|
||||
]
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# Show how to use Tally.get_values(...) with a CrossFilter and CrossScore\n",
|
||||
"fast_scatter_xs = fuel_xs.get_values(filters=['energy'], \n",
|
||||
|
|
@ -974,11 +1341,81 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 37,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
||||
" <th>cell</th>\n",
|
||||
" <th>energy [MeV]</th>\n",
|
||||
" <th>nuclide</th>\n",
|
||||
" <th>score</th>\n",
|
||||
" <th>mean</th>\n",
|
||||
" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
||||
" <tr>\n",
|
||||
" <th>0</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>U-238</td>\n",
|
||||
" <td>nu-fission</td>\n",
|
||||
" <td>0.000002</td>\n",
|
||||
" <td>1.284890e-08</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>1</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(0.0e+00 - 6.3e-07)</td>\n",
|
||||
" <td>U-235</td>\n",
|
||||
" <td>nu-fission</td>\n",
|
||||
" <td>0.867982</td>\n",
|
||||
" <td>7.022256e-03</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>2</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>U-238</td>\n",
|
||||
" <td>nu-fission</td>\n",
|
||||
" <td>0.082801</td>\n",
|
||||
" <td>6.087096e-04</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>3</th>\n",
|
||||
" <td>10000</td>\n",
|
||||
" <td>(6.3e-07 - 2.0e+01)</td>\n",
|
||||
" <td>U-235</td>\n",
|
||||
" <td>nu-fission</td>\n",
|
||||
" <td>0.093484</td>\n",
|
||||
" <td>5.275039e-04</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" cell energy [MeV] nuclide score mean std. dev.\n",
|
||||
"0 10000 (0.0e+00 - 6.3e-07) U-238 nu-fission 0.000002 1.284890e-08\n",
|
||||
"1 10000 (0.0e+00 - 6.3e-07) U-235 nu-fission 0.867982 7.022256e-03\n",
|
||||
"2 10000 (6.3e-07 - 2.0e+01) U-238 nu-fission 0.082801 6.087096e-04\n",
|
||||
"3 10000 (6.3e-07 - 2.0e+01) U-235 nu-fission 0.093484 5.275039e-04"
|
||||
]
|
||||
},
|
||||
"execution_count": 37,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# \"Slice\" the nu-fission data into a new derived Tally\n",
|
||||
"nu_fission_rates = fuel_rxn_rates.get_slice(scores=['nu-fission'])\n",
|
||||
|
|
@ -987,11 +1424,131 @@
|
|||
},
|
||||
{
|
||||
"cell_type": "code",
|
||||
"execution_count": null,
|
||||
"execution_count": 38,
|
||||
"metadata": {
|
||||
"collapsed": false
|
||||
},
|
||||
"outputs": [],
|
||||
"outputs": [
|
||||
{
|
||||
"data": {
|
||||
"text/html": [
|
||||
"<div style=\"max-height:1000px;max-width:1500px;overflow:auto;\">\n",
|
||||
"<table border=\"1\" class=\"dataframe\">\n",
|
||||
" <thead>\n",
|
||||
" <tr style=\"text-align: right;\">\n",
|
||||
" <th></th>\n",
|
||||
" <th>cell</th>\n",
|
||||
" <th>energy [MeV]</th>\n",
|
||||
" <th>nuclide</th>\n",
|
||||
" <th>score</th>\n",
|
||||
" <th>mean</th>\n",
|
||||
" <th>std. dev.</th>\n",
|
||||
" </tr>\n",
|
||||
" </thead>\n",
|
||||
" <tbody>\n",
|
||||
" <tr>\n",
|
||||
" <th>0</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.0e-08 - 1.1e-07)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>4.620525</td>\n",
|
||||
" <td>0.038249</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>1</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.1e-07 - 1.2e-06)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>2.036841</td>\n",
|
||||
" <td>0.013203</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>2</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.2e-06 - 1.3e-05)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>1.659916</td>\n",
|
||||
" <td>0.010107</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>3</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.3e-05 - 1.4e-04)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>1.861546</td>\n",
|
||||
" <td>0.013328</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>4</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.4e-04 - 1.5e-03)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>2.049664</td>\n",
|
||||
" <td>0.008215</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>5</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.5e-03 - 1.6e-02)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>2.162157</td>\n",
|
||||
" <td>0.010245</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>6</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.6e-02 - 1.7e-01)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>2.224496</td>\n",
|
||||
" <td>0.013796</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>7</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.7e-01 - 1.9e+00)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>1.997585</td>\n",
|
||||
" <td>0.009161</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
" <th>8</th>\n",
|
||||
" <td>10002</td>\n",
|
||||
" <td>(1.9e+00 - 2.0e+01)</td>\n",
|
||||
" <td>H-1</td>\n",
|
||||
" <td>scatter</td>\n",
|
||||
" <td>0.373472</td>\n",
|
||||
" <td>0.003922</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
"</table>\n",
|
||||
"</div>"
|
||||
],
|
||||
"text/plain": [
|
||||
" cell energy [MeV] nuclide score mean std. dev.\n",
|
||||
"0 10002 (1.0e-08 - 1.1e-07) H-1 scatter 4.620525 0.038249\n",
|
||||
"1 10002 (1.1e-07 - 1.2e-06) H-1 scatter 2.036841 0.013203\n",
|
||||
"2 10002 (1.2e-06 - 1.3e-05) H-1 scatter 1.659916 0.010107\n",
|
||||
"3 10002 (1.3e-05 - 1.4e-04) H-1 scatter 1.861546 0.013328\n",
|
||||
"4 10002 (1.4e-04 - 1.5e-03) H-1 scatter 2.049664 0.008215\n",
|
||||
"5 10002 (1.5e-03 - 1.6e-02) H-1 scatter 2.162157 0.010245\n",
|
||||
"6 10002 (1.6e-02 - 1.7e-01) H-1 scatter 2.224496 0.013796\n",
|
||||
"7 10002 (1.7e-01 - 1.9e+00) H-1 scatter 1.997585 0.009161\n",
|
||||
"8 10002 (1.9e+00 - 2.0e+01) H-1 scatter 0.373472 0.003922"
|
||||
]
|
||||
},
|
||||
"execution_count": 38,
|
||||
"metadata": {},
|
||||
"output_type": "execute_result"
|
||||
}
|
||||
],
|
||||
"source": [
|
||||
"# \"Slice\" the H-1 scatter data in the moderator Cell into a new derived Tally\n",
|
||||
"need_to_slice = sp.get_tally(name='need-to-slice')\n",
|
||||
|
|
|
|||
|
|
@ -1119,14 +1119,22 @@ class MultiGroupXS(object):
|
|||
|
||||
|
||||
class TotalXS(MultiGroupXS):
|
||||
"""A total multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(TotalXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(TotalXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'total'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'total' reaction rates in the spatial domain and energy groups
|
||||
of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'total']
|
||||
|
|
@ -1143,21 +1151,30 @@ class TotalXS(MultiGroupXS):
|
|||
|
||||
def compute_xs(self):
|
||||
"""Computes the multi-group total cross sections using OpenMC
|
||||
tally arithmetic."""
|
||||
tally arithmetic.
|
||||
"""
|
||||
|
||||
self._xs_tally = self.tallies['total'] / self.tallies['flux']
|
||||
super(TotalXS, self).compute_xs()
|
||||
|
||||
|
||||
class TransportXS(MultiGroupXS):
|
||||
"""A transport-corrected total multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(TransportXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(TransportXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'transport'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs three analog tallies to compute the 'flux',
|
||||
'total' and 'scatter-P1' reaction rates in the spatial domain and
|
||||
energy groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'total', 'scatter-P1']
|
||||
|
|
@ -1171,10 +1188,35 @@ class TransportXS(MultiGroupXS):
|
|||
filters = [[energy_filter], [energy_filter], [energyout_filter]]
|
||||
|
||||
# Initialize the Tallies
|
||||
super(TransportXS, self).create_tallies(scores, filters, keys, estimator)
|
||||
super(TransportXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
def load_from_statepoint(self, statepoint):
|
||||
"""Extracts tallies in an OpenMC StatePoint with the data needed to
|
||||
compute multi-group cross sections.
|
||||
|
||||
This method is needed to compute cross section data from tallies
|
||||
in an OpenMC StatePoint object.
|
||||
|
||||
NOTE: The statepoint must first be linked with an OpenMC Summary object.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
statepoint : openmc.StatePoint
|
||||
An OpenMC StatePoint object with tally data
|
||||
|
||||
Raises
|
||||
------
|
||||
ValueError
|
||||
When this method is called with a statepoint that has not been
|
||||
linked with a summary object.
|
||||
|
||||
"""
|
||||
|
||||
# Load the tallies from the statepoint using the parent class method
|
||||
super(TransportXS, self).load_from_statepoint(statepoint)
|
||||
|
||||
# Use tally slicing to remove scatter-P0 data from scatter-P1 tally
|
||||
scatter_p1 = self.tallies['scatter-P1']
|
||||
self.tallies['scatter-P1'] = scatter_p1.get_slice(scores=['scatter-P1'])
|
||||
self.tallies['scatter-P1'].filters[-1].type = 'energy'
|
||||
|
|
@ -1189,14 +1231,22 @@ class TransportXS(MultiGroupXS):
|
|||
|
||||
|
||||
class AbsorptionXS(MultiGroupXS):
|
||||
"""An absorption multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(AbsorptionXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(AbsorptionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'absorption'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'absorption' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'absorption']
|
||||
|
|
@ -1209,7 +1259,8 @@ class AbsorptionXS(MultiGroupXS):
|
|||
filters = [[energy_filter], [energy_filter]]
|
||||
|
||||
# Initialize the Tallies
|
||||
super(AbsorptionXS, self).create_tallies(scores, filters, keys, estimator)
|
||||
super(AbsorptionXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
def compute_xs(self):
|
||||
"""Computes the multi-group absorption cross sections using OpenMC
|
||||
|
|
@ -1220,14 +1271,22 @@ class AbsorptionXS(MultiGroupXS):
|
|||
|
||||
|
||||
class CaptureXS(MultiGroupXS):
|
||||
"""A capture multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(CaptureXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(CaptureXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'capture'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'capture' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'absorption', 'fission']
|
||||
|
|
@ -1252,14 +1311,22 @@ class CaptureXS(MultiGroupXS):
|
|||
|
||||
|
||||
class FissionXS(MultiGroupXS):
|
||||
"""A fission multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(FissionXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(FissionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'fission'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'fission' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'fission']
|
||||
|
|
@ -1283,14 +1350,22 @@ class FissionXS(MultiGroupXS):
|
|||
|
||||
|
||||
class NuFissionXS(MultiGroupXS):
|
||||
"""A fission production multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(NuFissionXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(NuFissionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'nu-fission'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'nu-fission' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'nu-fission']
|
||||
|
|
@ -1303,7 +1378,8 @@ class NuFissionXS(MultiGroupXS):
|
|||
filters = [[energy_filter], [energy_filter]]
|
||||
|
||||
# Initialize the Tallies
|
||||
super(NuFissionXS, self).create_tallies(scores, filters, keys, estimator)
|
||||
super(NuFissionXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
def compute_xs(self):
|
||||
"""Computes the multi-group nu-fission cross sections using OpenMC
|
||||
|
|
@ -1314,14 +1390,22 @@ class NuFissionXS(MultiGroupXS):
|
|||
|
||||
|
||||
class ScatterXS(MultiGroupXS):
|
||||
"""A scatter multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(ScatterXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(ScatterXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'scatter'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two tracklength tallies to compute the 'flux'
|
||||
and 'scatter' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'scatter']
|
||||
|
|
@ -1345,14 +1429,22 @@ class ScatterXS(MultiGroupXS):
|
|||
|
||||
|
||||
class NuScatterXS(MultiGroupXS):
|
||||
"""A nu-scatter multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(NuScatterXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(NuScatterXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'nu-scatter'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two analog tallies to compute the 'flux'
|
||||
and 'nu-scatter' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'nu-scatter']
|
||||
|
|
@ -1365,7 +1457,8 @@ class NuScatterXS(MultiGroupXS):
|
|||
filters = [[energy_filter], [energy_filter]]
|
||||
|
||||
# Initialize the Tallies
|
||||
super(NuScatterXS, self).create_tallies(scores, filters, keys, estimator)
|
||||
super(NuScatterXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
def compute_xs(self):
|
||||
"""Computes the nu-scattering multi-group cross section using OpenMC
|
||||
|
|
@ -1376,14 +1469,22 @@ class NuScatterXS(MultiGroupXS):
|
|||
|
||||
|
||||
class ScatterMatrixXS(MultiGroupXS):
|
||||
"""A scattering matrix multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(ScatterMatrixXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(ScatterMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'scatter matrix'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs three analog tallies to compute the 'flux',
|
||||
'scatter' and 'scatter-P1' reaction rates in the spatial domain and
|
||||
energy groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energy = openmc.Filter('energy', group_edges)
|
||||
|
|
@ -1397,7 +1498,8 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
keys = scores
|
||||
|
||||
# Initialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters, keys, estimator)
|
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super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
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keys, estimator)
|
||||
|
||||
def compute_xs(self, correction='P0'):
|
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"""Computes the multi-group scattering matrix using OpenMC
|
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|
|
@ -1425,9 +1527,9 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
self._xs_tally = rxn_tally / self.tallies['flux']
|
||||
super(ScatterMatrixXS, self).compute_xs()
|
||||
|
||||
def get_xs(self, in_groups='all', out_groups='all', subdomains='all',
|
||||
nuclides='all', order_groups='increasing',
|
||||
xs_type='macro', value='mean'):
|
||||
def get_xs(self, in_groups='all', out_groups='all',
|
||||
subdomains='all', nuclides='all', xs_type='macro',
|
||||
order_groups='increasing', value='mean'):
|
||||
"""Returns an array of multi-group cross sections.
|
||||
|
||||
This method constructs a 2D NumPy array for the requested scattering
|
||||
|
|
@ -1448,15 +1550,16 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
return the cross section summed over all nuclides.
|
||||
xs_type: {'macro' or 'micro'}
|
||||
Return the macro or micro cross section in units of cm^-1 or barns
|
||||
xs_type: {'macro' or 'micro'}
|
||||
Return the macro or micro cross section in units of cm^-1 or barns
|
||||
order_groups: {'increasing', 'decreasing'}
|
||||
Return the cross section indexed according to increasing (default)
|
||||
or decreasing energy groups (decreasing or increasing energies)
|
||||
value : str
|
||||
A string for the type of value to return - 'mean' (default),
|
||||
'std_dev' or 'rel_err' are accepted
|
||||
|
||||
Returns
|
||||
-------
|
||||
xs : ndarray
|
||||
ndarray
|
||||
A NumPy array of the multi-group cross section indexed in the order
|
||||
each group and subdomain is listed in the parameters.
|
||||
|
||||
|
|
@ -1500,8 +1603,6 @@ class ScatterMatrixXS(MultiGroupXS):
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filter_bins.append((self.energy_groups.get_group_bounds(group),))
|
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|
||||
# Construct a collection of the nuclides to retrieve from the xs tally
|
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# NOTE: We must not override the "nuclides" parameter since it is used
|
||||
# to retrieve atomic number densities for micro xs
|
||||
if self.by_nuclide:
|
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if nuclides == 'all' or nuclides == 'sum' or nuclides == ['sum']:
|
||||
query_nuclides = self.get_all_nuclides()
|
||||
|
|
@ -1516,8 +1617,9 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
xs = xs_tally.get_values(filters=filters,
|
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filter_bins=filter_bins, value=value)
|
||||
else:
|
||||
xs = self.xs_tally.get_values(filters=filters, filter_bins=filter_bins,
|
||||
nuclides=query_nuclides, value=value)
|
||||
xs = self.xs_tally.get_values(filters=filters,
|
||||
filter_bins=filter_bins,
|
||||
nuclides=query_nuclides, value=value)
|
||||
|
||||
xs = np.nan_to_num(xs)
|
||||
|
||||
|
|
@ -1533,7 +1635,6 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
# Reverse data if user requested increasing energy groups since
|
||||
# tally data is stored in order of increasing energies
|
||||
if order_groups == 'increasing':
|
||||
# Reshape tally data array with separate axes for domain and energy
|
||||
if in_groups == 'all':
|
||||
num_in_groups = self.num_groups
|
||||
else:
|
||||
|
|
@ -1542,6 +1643,8 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
num_out_groups = self.num_groups
|
||||
else:
|
||||
num_out_groups = len(out_groups)
|
||||
|
||||
# Reshape tally data array with separate axes for domain and energy
|
||||
num_subdomains = xs.shape[0] / (num_in_groups * num_out_groups)
|
||||
new_shape = (num_subdomains, num_in_groups, num_out_groups)
|
||||
new_shape += xs.shape[1:]
|
||||
|
|
@ -1648,13 +1751,15 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
for out_group in range(1, self.num_groups+1):
|
||||
string += template.format('', in_group, out_group)
|
||||
average = \
|
||||
self.get_xs([in_group], [out_group], [subdomain],
|
||||
[nuclide], xs_type=xs_type, value='mean')
|
||||
self.get_xs([in_group], [out_group],
|
||||
[subdomain], [nuclide],
|
||||
xs_type=xs_type, value='mean')
|
||||
rel_err = \
|
||||
self.get_xs([in_group], [out_group], [subdomain],
|
||||
[nuclide], xs_type=xs_type, value='rel_err') * 100
|
||||
average = np.nan_to_num(average.flatten())[0]
|
||||
rel_err = np.nan_to_num(rel_err.flatten())[0]
|
||||
self.get_xs([in_group], [out_group],
|
||||
[subdomain], [nuclide],
|
||||
xs_type=xs_type, value='rel_err')
|
||||
average = average.flatten()[0]
|
||||
rel_err = rel_err.flatten()[0] * 100.
|
||||
string += '{:1.2e} +/- {:1.2e}%'.format(average, rel_err)
|
||||
string += '\n'
|
||||
string += '\n'
|
||||
|
|
@ -1665,14 +1770,22 @@ class ScatterMatrixXS(MultiGroupXS):
|
|||
|
||||
|
||||
class NuScatterMatrixXS(ScatterMatrixXS):
|
||||
"""A scattering production matrix multi-group cross section."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
super(NuScatterMatrixXS, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
super(NuScatterMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'nu-scatter matrix'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs three analog tallies to compute the 'flux',
|
||||
'nu-scatter' and 'scatter-P1' reaction rates in the spatial domain and
|
||||
energy groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'scatter', 'scatter-P1']
|
||||
|
|
@ -1686,9 +1799,11 @@ class NuScatterMatrixXS(ScatterMatrixXS):
|
|||
filters = [[energy], [energy, energyout], [energyout]]
|
||||
|
||||
# Intialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters, keys, estimator)
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
class Chi(MultiGroupXS):
|
||||
"""The fission spectrum."""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name=''):
|
||||
|
|
@ -1696,7 +1811,13 @@ class Chi(MultiGroupXS):
|
|||
self._rxn_type = 'chi'
|
||||
|
||||
def create_tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section."""
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two analog tallies to compute 'nu-fission'
|
||||
reaction rates with 'energy' and 'energyout' filters in the spatial
|
||||
domain and energy groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['nu-fission', 'nu-fission']
|
||||
|
|
@ -1732,8 +1853,8 @@ class Chi(MultiGroupXS):
|
|||
super(Chi, self).compute_xs()
|
||||
|
||||
def get_xs(self, groups='all', subdomains='all', nuclides='all',
|
||||
order_groups='increasing', xs_type='macro', value='mean'):
|
||||
"""Returns an array of multi-group cross sections.
|
||||
xs_type='macro', order_groups='increasing', value='mean'):
|
||||
"""Returns an array of the fission spectrum.
|
||||
|
||||
This method constructs a 2D NumPy array for the requested multi-group
|
||||
cross section data data for one or more energy groups and subdomains.
|
||||
|
|
@ -1749,18 +1870,19 @@ class Chi(MultiGroupXS):
|
|||
special string 'all' (default) will return the cross sections for
|
||||
all nuclides in the spatial domain. The special string 'sum' will
|
||||
return the cross section summed over all nuclides.
|
||||
xs_type: {'macro' or 'micro'}
|
||||
Return the macro or micro cross section in units of cm^-1 or barns
|
||||
xs_type: {'macro' or 'micro'}
|
||||
This parameter is not relevant for chi but is included here to
|
||||
mirror the parent MultiGroupXS.get_xs(...) class method
|
||||
order_groups: {'increasing', 'decreasing'}
|
||||
Return the cross section indexed according to increasing (default)
|
||||
or decreasing energy groups (decreasing or increasing energies)
|
||||
value : str
|
||||
A string for the type of value to return - 'mean' (default),
|
||||
'std_dev' or 'rel_err' are accepted
|
||||
|
||||
Returns
|
||||
-------
|
||||
xs : ndarray
|
||||
ndarray
|
||||
A NumPy array of the multi-group cross section indexed in the order
|
||||
each group, subdomain and nuclide is listed in the parameters.
|
||||
|
||||
|
|
@ -1812,8 +1934,8 @@ class Chi(MultiGroupXS):
|
|||
nu_fission_in = nu_fission_in.summation(nuclides=nuclides)
|
||||
nu_fission_out = nu_fission_out.summation(nuclides=nuclides)
|
||||
|
||||
# Compute chi and store it as the xs_tally attribute so we can use
|
||||
# the generic get_xs(...) method
|
||||
# Compute chi and store it as the xs_tally attribute so we can
|
||||
# use the generic get_xs(...) method
|
||||
xs_tally = nu_fission_out / nu_fission_in
|
||||
xs = xs_tally.get_values(filters=filters,
|
||||
filter_bins=filter_bins, value=value)
|
||||
|
|
@ -1821,13 +1943,15 @@ class Chi(MultiGroupXS):
|
|||
# Get chi for all nuclides in the domain
|
||||
elif nuclides == 'all':
|
||||
nuclides = self.get_all_nuclides()
|
||||
xs = self.xs_tally.get_values(filters=filters, filter_bins=filter_bins,
|
||||
xs = self.xs_tally.get_values(filters=filters,
|
||||
filter_bins=filter_bins,
|
||||
nuclides=nuclides, value=value)
|
||||
|
||||
# Get chi for user-specified nuclides in the domain
|
||||
else:
|
||||
cv.check_iterable_type('nuclides', nuclides, basestring)
|
||||
xs = self.xs_tally.get_values(filters=filters, filter_bins=filter_bins,
|
||||
xs = self.xs_tally.get_values(filters=filters,
|
||||
filter_bins=filter_bins,
|
||||
nuclides=nuclides, value=value)
|
||||
|
||||
# If chi was computed as an average of nuclides in the domain
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue