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4
docs/source/conf.py
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@ -55,7 +55,7 @@ copyright = u'2011-2015, Massachusetts Institute of Technology'
# The short X.Y version.
version = "0.7"
# The full version, including alpha/beta/rc tags.
release = "0.7.0"
release = "0.7.1"
# The language for content autogenerated by Sphinx. Refer to documentation
# for a list of supported languages.
@ -200,7 +200,7 @@ latex_elements = {
\usepackage{enumitem}
\usepackage{amsfonts}
\usepackage{amsmath}
\setlistdepth{9}
\setlistdepth{99}
\usepackage{tikz}
\usetikzlibrary{shapes,snakes,shadows,arrows,calc,decorations.markings,patterns,fit,matrix,spy}
\usepackage{fixltx2e}

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@ -26,6 +26,10 @@ Overviews
Benchmarking
------------
- Khurrum S. Chaudri and Sikander M. Mirza, "Burnup dependent Monte Carlo
neutron physics calculations of IAEA MTR benchmark," *Prog. Nucl. Energy*,
**81**, 43-52 (2015). `<http://dx.doi.org/j.pnucene.2014.12.018>`_
- Daniel J. Kelly, Brian N. Aviles, Paul K. Romano, Bryan R. Herman,
Nicholas E. Horelik, and Benoit Forget, "Analysis of select BEAVRS PWR
benchmark cycle 1 results using MC21 and OpenMC," *Proc. PHYSOR*, Kyoto,
@ -57,13 +61,8 @@ Coupling and Multi-physics
- Bryan R. Herman, Benoit Forget, and Kord Smith, "Progress toward Monte
Carlo-thermal hydraulic coupling using low-order nonlinear diffusion
acceleration methods." In press, *Ann. Nucl. Energy*,
(2014). `<http://dx.doi.org/10.1016/j.anucene/2014.10.029>`_
- Adam G. Nelson and William R. Martin, "Improved Convergence of Monte Carlo
Generated Multi-Group Scattering Moments," *Proc. Int. Conf. Mathematics and
Computational Methods Applied to Nuclear Science and Engineering*, Sun Valley,
Idaho, May 5--9 (2013).
acceleration methods." *Ann. Nucl. Energy*, **84**, 63-72
(2015). `<http://dx.doi.org/10.1016/j.anucene.2014.10.029>`_
- Bryan R. Herman, Benoit Forget, and Kord Smith, "Utilizing CMFD in OpenMC to
Estimate Dominance Ratio and Adjoint," *Trans. Am. Nucl. Soc.*, **109**,
@ -81,19 +80,65 @@ Geometry
Miscellaneous
-------------
- William Boyd, Sterling Harper, and Paul K. Romano, "Equipping OpenMC for the
big data era," Accepted, *PHYSOR 2016*, Sun Valley, Idaho, May 1-5, 2016.
- Qicang Shen, William Boyd, Benoit Forget, and Kord Smith, "Tally precision
triggers for the OpenMC Monte Carlo code," *Trans. Am. Nucl. Soc.*, **112**,
637-640 (2015).
- Timothy P. Burke, Brian C. Kiedrowski, and William R. Martin, "Flux and
Reaction Rate Kernel Density Estimators in OpenMC," *Trans. Am. Nucl. Soc.*,
**109**, 683-686 (2013).
------------------------------------
Multi-group Cross Section Generation
------------------------------------
- Adam G. Nelson and William R. Martin, "Improved Monte Carlo tallying of
multi-group scattering moments using the NDPP code," *Trans. Am. Nucl. Soc.*,
**113**, 645-648 (2015)
- Adam G. Nelson and William R. Martin, "Improved Monte Carlo tallying of
multi-group scattering moment matrices," *Trans. Am. Nucl. Soc.*, **110**,
217-220 (2014).
- Adam G. Nelson and William R. Martin, "Improved Convergence of Monte Carlo
Generated Multi-Group Scattering Moments," *Proc. Int. Conf. Mathematics and
Computational Methods Applied to Nuclear Science and Engineering*, Sun Valley,
Idaho, May 5--9 (2013).
------------
Nuclear Data
------------
- Colin Josey, Pablo Ducru, Benoit Forget, and Kord Smith, "Windowed multipole
for cross section Doppler broadening," *J. Comput. Phys.*, In Press
(2016). `<http://dx.doi.org/10.1016/jcp.2015.08.013>`_
- Colin Josey, Benoit Forget, and Kord Smith, "Windowed multipole sensitivity to
target accuracy of the optimization procedure," *J. Nucl. Sci. Technol.*,
**52**, 987-992 (2015). `<http://dx.doi.org/10.1080/00223131.2015.1035353>`_
- Jonathan A. Walsh, Paul K. Romano, Benoit Forget, and Kord S. Smith,
"Optimizations of the energy grid search algorithm in continuous-energy Monte
Carlo particle transport codes", *Comput. Phys. Commun.*, **196**, 134-142
(2015). `<http://dx.doi.org/10.1016/j.cpc.2015.05.025>`_
- Jonathan A. Walsh, Benoit Forget, Kord S. Smith, Brian C. Kiedrowski, and
Forrest B. Brown, "Direct, on-the-fly calculation of unresolved resonance
region cross sections in Monte Carlo simulations," *Proc. Joint
Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
- Amanda L. Lund, Andrew R. Siegel, Benoit Forget, Colin Josey, and
Paul K. Romano, "Using fractional cascading to accelerate cross section
lookups in Monte Carlo particle transport calculations," *Proc. Joint
Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
- Ronald O. Rahaman, Andrew R. Siegel, and Paul K. Romano, "Monte Carlo
performance analysis for varying cross section parameter regimes,"
*Proc. Joint Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
- Paul K. Romano and Timothy H. Trumbull, "Comparison of algorithms for Doppler
broadening pointwise tabulated cross sections," *Ann. Nucl. Energy*, **75**,
358--364 (2015). `<http://dx.doi.org/10.1016/j.anucene.2014.08.046>`_
@ -114,6 +159,10 @@ Nuclear Data
Parallelism
-----------
- Paul K. Romano, John R. Tramm, and Andrew R. Siegel, "Efficacy of hardware
threading for Monte Carlo particle transport calculations on multi- and
many-core systems," Accepted, *PHYSOR 2016*, Sun Valley, Idaho, May 1-5, 2016.
- David Ozog, Allen D. Malony, and Andrew R. Siegel, "A performance analysis of
SIMD algorithms for Monte Carlo simulations of nuclear reactor cores,"
*Proc. IEEE Int. Parallel and Distributed Processing Symposium*, Hyderabad,

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.. _notebook_mgxs_part_i:
=========================
MGXS Part I: Introduction
=========================
.. only:: html
.. notebook:: mgxs-part-i.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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.. _notebook_mgxs_part_ii:
===============================
MGXS Part II: Advanced Features
===============================
.. only:: html
.. notebook:: mgxs-part-ii.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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.. _notebook_mgxs_part_iii:
========================
MGXS Part III: Libraries
========================
.. only:: html
.. notebook:: mgxs-part-iii.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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====================================
Multi-Group Cross Section Generation
====================================
.. only:: html
.. notebook:: multi-group-cross-sections.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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@ -126,7 +126,7 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"Now let's move on to the geometry. This problem will be a square array of fuel pins, which we can use OpenMC's lattice/universe feature for. The basic universe will have three regions for the fuel, the clad, and the surrounding coolant. The first step is to create the bounding surfaces for fuel and clad, as well as the outer bounding surfaces of the problem."
"Now let's move on to the geometry. This problem will be a square array of fuel pins for which we can use OpenMC's lattice/universe feature. The basic universe will have three regions for the fuel, the clad, and the surrounding coolant. The first step is to create the bounding surfaces for fuel and clad, as well as the outer bounding surfaces of the problem."
]
},
{
@ -155,7 +155,7 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"With the surfaces defined, we can now create cells that are defined by intersections of half-spaces created by the surfaces."
"With the surfaces defined, we can now construct a fuel pin cell from cells that are defined by intersections of half-spaces created by the surfaces."
]
},
{
@ -192,7 +192,7 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"Using the pin cell universe, we can construct a 17x17 rectangular lattice with a 1.26cm pitch."
"Using the pin cell universe, we can construct a 17x17 rectangular lattice with a 1.26 cm pitch."
]
},
{
@ -242,7 +242,7 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"We now must create a geometry that is assigned a root universe, put the geometry into a geometry file, and export it to XML."
"We now must create a geometry that is assigned a root universe, put the geometry into a `GeometryFile` object, and export it to XML."
]
},
{

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@ -363,7 +363,26 @@
"outputs": [
{
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@ -573,8 +592,8 @@
" Copyright: 2011-2015 Massachusetts Institute of Technology\n",
" License: http://mit-crpg.github.io/openmc/license.html\n",
" Version: 0.7.0\n",
" Git SHA1: 21738db07debeabde824c9b955bd3bf0c9a16366\n",
" Date/Time: 2015-10-28 21:15:07\n",
" Git SHA1: 74ffcb447521c968fb64fdaa63e40598783f2fba\n",
" Date/Time: 2015-11-25 14:20:51\n",
" MPI Processes: 1\n",
"\n",
" ===========================================================================\n",
@ -615,13 +634,13 @@
" 11/1 1.07867 1.05536 +/- 0.01277\n",
" 12/1 1.04203 1.05345 +/- 0.01096\n",
" 13/1 1.04482 1.05237 +/- 0.00955\n",
" 14/1 1.04117 1.05113 +/- 0.00852\n",
" 15/1 1.07581 1.05360 +/- 0.00801\n",
" 16/1 1.04235 1.05257 +/- 0.00731\n",
" 17/1 1.02710 1.05045 +/- 0.00701\n",
" 18/1 1.01970 1.04809 +/- 0.00687\n",
" 19/1 1.01022 1.04538 +/- 0.00691\n",
" 20/1 1.01449 1.04332 +/- 0.00675\n",
" 14/1 1.04116 1.05113 +/- 0.00852\n",
" 15/1 1.07569 1.05358 +/- 0.00800\n",
" 16/1 1.04188 1.05252 +/- 0.00732\n",
" 17/1 1.03775 1.05129 +/- 0.00679\n",
" 18/1 0.98462 1.04616 +/- 0.00808\n",
" 19/1 1.08613 1.04902 +/- 0.00801\n",
" 20/1 1.00571 1.04613 +/- 0.00800\n",
" Creating state point statepoint.20.h5...\n",
"\n",
" ===========================================================================\n",
@ -631,27 +650,27 @@
"\n",
" =======================> TIMING STATISTICS <=======================\n",
"\n",
" Total time for initialization = 6.3800E-01 seconds\n",
" Reading cross sections = 1.3500E-01 seconds\n",
" Total time in simulation = 2.3556E+01 seconds\n",
" Time in transport only = 2.3532E+01 seconds\n",
" Time in inactive batches = 3.1100E+00 seconds\n",
" Time in active batches = 2.0446E+01 seconds\n",
" Total time for initialization = 7.9600E-01 seconds\n",
" Reading cross sections = 2.1200E-01 seconds\n",
" Total time in simulation = 1.8740E+01 seconds\n",
" Time in transport only = 1.8727E+01 seconds\n",
" Time in inactive batches = 2.5970E+00 seconds\n",
" Time in active batches = 1.6143E+01 seconds\n",
" Time synchronizing fission bank = 2.0000E-03 seconds\n",
" Sampling source sites = 1.0000E-03 seconds\n",
" SEND/RECV source sites = 1.0000E-03 seconds\n",
" Time accumulating tallies = 1.0000E-03 seconds\n",
" Total time for finalization = 3.0000E-03 seconds\n",
" Total time elapsed = 2.4210E+01 seconds\n",
" Calculation Rate (inactive) = 4019.29 neutrons/second\n",
" Calculation Rate (active) = 1834.10 neutrons/second\n",
" Time accumulating tallies = 0.0000E+00 seconds\n",
" Total time for finalization = 2.0000E-03 seconds\n",
" Total time elapsed = 1.9553E+01 seconds\n",
" Calculation Rate (inactive) = 4813.25 neutrons/second\n",
" Calculation Rate (active) = 2322.99 neutrons/second\n",
"\n",
" ============================> RESULTS <============================\n",
"\n",
" k-effective (Collision) = 1.03935 +/- 0.00682\n",
" k-effective (Track-length) = 1.04332 +/- 0.00675\n",
" k-effective (Absorption) = 1.03845 +/- 0.00598\n",
" Combined k-effective = 1.04024 +/- 0.00523\n",
" k-effective (Collision) = 1.04597 +/- 0.00663\n",
" k-effective (Track-length) = 1.04613 +/- 0.00800\n",
" k-effective (Absorption) = 1.04087 +/- 0.00627\n",
" Combined k-effective = 1.04322 +/- 0.00570\n",
" Leakage Fraction = 0.00000 +/- 0.00000\n",
"\n"
]
@ -742,7 +761,7 @@
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@ -756,10 +775,10 @@
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@ -767,7 +786,7 @@
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},
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@ -802,7 +821,7 @@
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@ -817,19 +836,19 @@
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@ -862,11 +881,12 @@
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@ -876,18 +896,19 @@
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" <td>0.010602</td>\n",
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@ -921,7 +942,7 @@
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@ -937,12 +958,12 @@
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" <td>10000</td>\n",
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" <td>0.803413</td>\n",
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@ -950,7 +971,7 @@
],
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" energy [MeV] cell nuclide score mean std. dev.\n",
"0 (0.0e+00 - 6.2e-01) 10000 total absorption 0.803413 0.007031"
"0 (0.0e+00 - 6.2e-01) 10000 total absorption 0.803182 0.008664"
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"execution_count": 29,
@ -982,12 +1003,13 @@
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@ -997,19 +1019,23 @@
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" energy [MeV] nuclide score mean std. dev.\n",
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" energy [MeV] cell nuclide score mean \\\n",
"0 (0.0e+00 - 6.2e-01) 10000 total (nu-fission / absorption) 1.237982 \n",
"\n",
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"execution_count": 30,
@ -1040,12 +1066,13 @@
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@ -1055,22 +1082,23 @@
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" energy [MeV] nuclide \\\n",
"0 (0.0e+00 - 6.2e-01) total \n",
" energy [MeV] cell nuclide \\\n",
"0 (0.0e+00 - 6.2e-01) 10000 total \n",
"\n",
" score mean std. dev. \n",
"0 (((absorption * nu-fission) * absorption) * (n... 1.040166 0.019018 "
"0 (((absorption * nu-fission) * absorption) * (n... 1.040687 0.022989 "
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"execution_count": 31,
@ -1118,7 +1146,7 @@
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@ -1134,100 +1162,100 @@
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" <td> 1.024510e-03</td>\n",
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" <th>6</th>\n",
" <td>10000</td>\n",
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" <th>7</th>\n",
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" 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",
" cell energy [MeV] nuclide score mean \\\n",
"0 10000 (0.0e+00 - 6.3e-07) (U-238 / total) (nu-fission / flux) 0.000001 \n",
"1 10000 (0.0e+00 - 6.3e-07) (U-238 / total) (scatter / flux) 0.209990 \n",
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"3 10000 (0.0e+00 - 6.3e-07) (U-235 / total) (scatter / flux) 0.005555 \n",
"4 10000 (6.3e-07 - 2.0e+01) (U-238 / total) (nu-fission / flux) 0.007190 \n",
"5 10000 (6.3e-07 - 2.0e+01) (U-238 / total) (scatter / flux) 0.227843 \n",
"6 10000 (6.3e-07 - 2.0e+01) (U-235 / total) (nu-fission / flux) 0.008086 \n",
"7 10000 (6.3e-07 - 2.0e+01) (U-235 / total) (scatter / flux) 0.003365 \n",
"\n",
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"0 6.657029e-07 7.377419e-09 \n",
"1 2.099891e-01 2.303838e-03 \n",
"2 3.564204e-01 3.951669e-03 \n",
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"6 8.066738e-03 5.254797e-05 \n",
"7 3.366802e-03 1.647058e-05 "
" std. dev. \n",
"0 8.078651e-09 \n",
"1 2.449396e-03 \n",
"2 4.364366e-03 \n",
"3 6.495710e-05 \n",
"4 7.596666e-05 \n",
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@ -1258,11 +1286,11 @@
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@ -1290,9 +1318,9 @@
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@ -1314,8 +1342,8 @@
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@ -1344,7 +1372,7 @@
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@ -1400,10 +1428,10 @@
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@ -1443,84 +1471,84 @@
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" <td> H-1</td>\n",
" <td> scatter</td>\n",
" <td> 2.209947</td>\n",
" <td> 0.013848</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>2.011925</td>\n",
" <td>0.009406</td>\n",
" <td> 10002</td>\n",
" <td> (1.7e-01 - 1.9e+00)</td>\n",
" <td> H-1</td>\n",
" <td> scatter</td>\n",
" <td> 2.006967</td>\n",
" <td> 0.009368</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>0.371280</td>\n",
" <td>0.003949</td>\n",
" <td> 10002</td>\n",
" <td> (1.9e+00 - 2.0e+01)</td>\n",
" <td> H-1</td>\n",
" <td> scatter</td>\n",
" <td> 0.373895</td>\n",
" <td> 0.002964</td>\n",
" </tr>\n",
" </tbody>\n",
"</table>\n",
@ -1528,15 +1556,15 @@
],
"text/plain": [
" cell energy [MeV] nuclide score mean std. dev.\n",
"0 10002 (1.0e-08 - 1.1e-07) H-1 scatter 4.619398 0.040124\n",
"1 10002 (1.1e-07 - 1.2e-06) H-1 scatter 2.030757 0.011239\n",
"2 10002 (1.2e-06 - 1.3e-05) H-1 scatter 1.658488 0.009777\n",
"3 10002 (1.3e-05 - 1.4e-04) H-1 scatter 1.853002 0.007378\n",
"4 10002 (1.4e-04 - 1.5e-03) H-1 scatter 2.050773 0.012484\n",
"5 10002 (1.5e-03 - 1.6e-02) H-1 scatter 2.131759 0.007821\n",
"6 10002 (1.6e-02 - 1.7e-01) H-1 scatter 2.213710 0.015159\n",
"7 10002 (1.7e-01 - 1.9e+00) H-1 scatter 2.011925 0.009406\n",
"8 10002 (1.9e+00 - 2.0e+01) H-1 scatter 0.371280 0.003949"
"0 10002 (1.0e-08 - 1.1e-07) H-1 scatter 4.630154 0.044512\n",
"1 10002 (1.1e-07 - 1.2e-06) H-1 scatter 2.042984 0.011429\n",
"2 10002 (1.2e-06 - 1.3e-05) H-1 scatter 1.657517 0.008617\n",
"3 10002 (1.3e-05 - 1.4e-04) H-1 scatter 1.863326 0.008848\n",
"4 10002 (1.4e-04 - 1.5e-03) H-1 scatter 2.043916 0.014195\n",
"5 10002 (1.5e-03 - 1.6e-02) H-1 scatter 2.134458 0.007561\n",
"6 10002 (1.6e-02 - 1.7e-01) H-1 scatter 2.209947 0.013848\n",
"7 10002 (1.7e-01 - 1.9e+00) H-1 scatter 2.006967 0.009368\n",
"8 10002 (1.9e+00 - 2.0e+01) H-1 scatter 0.373895 0.002964"
]
},
"execution_count": 38,
@ -1569,7 +1597,7 @@
"name": "python",
"nbconvert_exporter": "python",
"pygments_lexer": "ipython2",
"version": "2.7.6"
"version": "2.7.10"
}
},
"nbformat": 4,

4
docs/source/pythonapi/index.rst
View file

@ -74,7 +74,9 @@ on a given module or class.
examples/post-processing
examples/pandas-dataframes
examples/tally-arithmetic
examples/multi-group-cross-sections
examples/mgxs-part-i
examples/mgxs-part-ii
examples/mgxs-part-iii
.. _Jupyter: https://jupyter.org/
.. _NumPy: http://www.numpy.org/

84
docs/source/releasenotes.rst
View file

@ -1,9 +1,30 @@
.. _releasenotes:
==============================
Release Notes for OpenMC 0.7.0
Release Notes for OpenMC 0.7.1
==============================
This release of OpenMC provides some substantial improvements over version
0.7.0. Non-simple cell regions can now be defined through the ``|`` (union) and
``~`` (complement) operators. Similar changes in the Python API also allow
complex cell regions to be defined. A true secondary particle bank now exists;
this is crucial for photon transport (to be added in the next minor release). A
rich API for multi-group cross section generation has been added via the
``openmc.mgxs`` Python module.
Various improvements to tallies have also been made. It is now possible to
explicitly specify that a collision estimator be used in a tally. A new
``delayedgroup`` filter and ``delayed-nu-fission`` score allow a user to obtain
delayed fission neutron production rates filtered by delayed group. Finally, the
new ``inverse-velocity`` score may be useful for calculating kinetics
parameters.
.. caution:: In previous versions, depending on how OpenMC was compiled binary
output was either given in HDF5 or a flat binary format. With this
version, all binary output is now HDF5 which means you **must**
have HDF5 in order to install OpenMC. Please consult the user's
guide for instructions on how to compile with HDF5.
-------------------
System Requirements
-------------------
@ -17,36 +38,41 @@ the problem at hand (mostly on the number of nuclides in the problem).
New Features
------------
- Complete Python API
- Python 3 compatability for all scripts
- All scripts consistently named openmc-* and installed together
- New 'distribcell' tally filter for repeated cells
- Ability to specify outer lattice universe
- XML input validation utility (openmc-validate-xml)
- Support for hexagonal lattices
- Material union energy grid method
- Tally triggers
- Remove dependence on PETSc
- Significant OpenMP performance improvements
- Support for Fortran 2008 MPI interface
- Use of Travis CI for continuous integration
- Simplifications and improvements to test suite
- Support for complex cell regions (union and complement operators)
- Generic quadric surface type
- Improved handling of secondary particles
- Binary output is now solely HDF5
- ``openmc.mgxs`` Python module enabling multi-group cross section generation
- Collision estimator for tallies
- Delayed fission neutron production tallies with ability to filter by delayed
group
- Inverse velocity tally score
- Performance improvements for binary search
- Performance improvements for reaction rate tallies
---------
Bug Fixes
---------
- b5f712_: Fix bug in spherical harmonics tallies
- e6675b_: Ensure all constants are double precision
- 04e2c1_: Fix potential bug in sample_nuclide routine
- 6121d9_: Fix bugs related to particle track files
- 2f0e89_: Fixes for nuclide specification in tallies
- 299322_: Bug with material filter when void material present
- d74840_: Fix triggers on tallies with multiple filters
- c29a81_: Correctly handle maximum transport energy
- 3edc23_: Fixes in the nu-scatter score
- 629e3b_: Assume unspecified surface coefficients are zero in Python API
- 5dbe8b_: Fix energy filters for openmc-plot-mesh-tally
- ff66f4_: Fixes in the openmc-plot-mesh-tally script
- 441fd4_: Fix bug in kappa-fission score
- 7e5974_: Allow fixed source simulations from Python API
.. _b5f712: https://github.com/mit-crpg/openmc/commit/b5f712
.. _e6675b: https://github.com/mit-crpg/openmc/commit/e6675b
.. _04e2c1: https://github.com/mit-crpg/openmc/commit/04e2c1
.. _6121d9: https://github.com/mit-crpg/openmc/commit/6121d9
.. _2f0e89: https://github.com/mit-crpg/openmc/commit/2f0e89
.. _299322: https://github.com/mit-crpg/openmc/commit/299322
.. _d74840: https://github.com/mit-crpg/openmc/commit/d74840
.. _c29a81: https://github.com/mit-crpg/openmc/commit/c29a81
.. _3edc23: https://github.com/mit-crpg/openmc/commit/3edc23
.. _629e3b: https://github.com/mit-crpg/openmc/commit/629e3b
.. _5dbe8b: https://github.com/mit-crpg/openmc/commit/5dbe8b
.. _ff66f4: https://github.com/mit-crpg/openmc/commit/ff66f4
.. _441fd4: https://github.com/mit-crpg/openmc/commit/441fd4
.. _7e5974: https://github.com/mit-crpg/openmc/commit/7e5974
------------
Contributors
@ -55,13 +81,11 @@ Contributors
This release contains new contributions from the following people:
- `Will Boyd <wbinventor@gmail.com>`_
- `Matt Ellis <mellis13@mit.edu>`_
- `Sterling Harper <sterlingmharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Nicholas Horelik <nicholas.horelik@gmail.com>`_
- `Bryan Herman <hermab53@gmail.com>`_
- `Colin Josey <cjosey@mit.edu>`_
- `William Lyu <PaleNeutron@users.noreply.github.com>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Anthony Scopatz <scopatz@gmail.com>`_
- `Kelly Rowland <kellylynnerowland@gmail.com>`_
- `Sam Shaner <samuelshaner@gmail.com>`_
- `Jon Walsh <walshjon@mit.edu>`_

13
docs/source/usersguide/input.rst
View file

@ -1147,10 +1147,12 @@ Each ``material`` element can have the following attributes or sub-elements:
:density:
An element with attributes/sub-elements called ``value`` and ``units``. The
``value`` attribute is the numeric value of the density while the ``units``
can be "g/cm3", "kg/m3", "atom/b-cm", "atom/cm3", "sum", or "macro".
The "sum" unit indicates that the density should be calculated as the sum
of the atom fractions for each nuclide in the material. This should not be
used in conjunction with weight percents. The "macro" unit is used with
can be "g/cm3", "kg/m3", "atom/b-cm", "atom/cm3", or "sum". The "sum" unit
indicates that values appearing in ``ao`` attributes for ``<nuclide>`` and
``<element>`` sub-elements are to be interpreted as nuclide/element
densities in atom/b-cm, and the total density of the material is taken as
the sum of all nuclides/elements. The "sum" option cannot be used in
conjunction with weight percents. The "macro" unit is used with
a ``macroscopic`` quantity to indicate that the density is already included
in the library and thus not needed here. However, if a value is provided
for the ``value``, then this is treated as a number density multiplier on
@ -1964,9 +1966,6 @@ attributes/sub-elements:
automatically assumes a one energy group calculation over the entire
energy range.
.. note:: When running in the multi-group :ref:`energy_mode`, the
energy bins must match the data library's group boundaries.
:albedo:
Surface ratio of incoming to outgoing partial currents on global boundary
conditions. They are listed in the following order: -x +x -y +y -z +z.

11
docs/source/usersguide/install.rst
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@ -8,7 +8,7 @@ Installation and Configuration
Installing on Ubuntu with PPA
-----------------------------
For users with Ubuntu 11.10 or later, a binary package for OpenMC is available
For users with Ubuntu 15.04 or later, a binary package for OpenMC is available
through a Personal Package Archive (PPA) and can be installed through the APT
package manager. First, add the following PPA to the repository sources:
@ -28,6 +28,9 @@ Now OpenMC should be recognized within the repository and can be installed:
sudo apt-get install openmc
Binary packages from this PPA may exist for earlier versions of Ubuntu, but they
are no longer supported.
--------------------
Building from Source
--------------------
@ -74,6 +77,12 @@ Prerequisites
You may omit ``--enable-parallel`` if you want to compile HDF5_ in serial.
.. important::
OpenMC uses various parts of the HDF5 Fortran 2003 API; as such you
must include ``--enable-fortran2003`` or else OpenMC will not be able
to compile.
On Debian derivatives, HDF5 and/or parallel HDF5 can be installed through
the APT package manager: