Documentation for version 0.7.1.

_sources/devguide/index.txt

@@ -16,6 +16,4 @@ as debugging.
     styleguide
     workflow
     xml-parsing
-    statepoint
-    voxel
     docbuild

_sources/devguide/statepoint.txt

@@ -1,291 +0,0 @@
-.. _devguide_statepoint:
-
-======================================
-State Point Binary File Specifications
-======================================
-
-The current revision of the statepoint binary file is 13.
-
-**integer(4) FILETYPE_STATEPOINT**
-
-    Flags whether this file is a statepoint file or a particle restart file.
-
-**integer(4) REVISION_STATEPOINT**
-
-    Revision of the binary state point file. Any time a change is made in the
-    format of the state-point file, this integer is incremented.
-
-**integer(4) VERSION_MAJOR**
-
-    Major version number for OpenMC
-
-**integer(4) VERSION_MINOR**
-
-    Minor version number for OpenMC
-
-**integer(4) VERSION_RELEASE**
-
-    Release version number for OpenMC
-
-**character(19) time_stamp**
-
-    Date and time the state point was written.
-
-**character(255) path**
-
-    Absolute path to directory containing input files.
-
-**integer(8) seed**
-
-    Pseudo-random number generator seed.
-
-**integer(4) run_mode**
-
-    run mode used. The modes are described in constants.F90.
-
-**integer(8) n_particles**
-
-    Number of particles used per generation.
-
-**integer(4) current_batch**
-
-    The number of batches already simulated.
-
-if (run_mode == MODE_EIGENVALUE)
-
-    **integer(4) n_inactive**
-
-        Number of inactive batches
-
-    **integer(4) gen_per_batch**
-
-        Number of generations per batch for criticality calculations
-
-    *do i = 1, current_batch \* gen_per_batch*
-
-        **real(8) k_generation(i)**
-
-             k-effective for the i-th total generation
-
-    *do i = 1, current_batch \* gen_per_batch*
-
-        **real(8) entropy(i)**
-
-            Shannon entropy for the i-th total generation
-
-    **real(8) k_col_abs**
-
-        Sum of product of collision/absorption estimates of k-effective
-
-    **real(8) k_col_tra**
-
-        Sum of product of collision/track-length estimates of k-effective
-
-    **real(8) k_abs_tra**
-
-        Sum of product of absorption/track-length estimates of k-effective
-
-    **real(8) k_combined(2)**
-
-        Mean and standard deviation of a combined estimate of k-effective
-
-    **integer(4) cmfd_on**
-
-        Flag that cmfd is on
-
-    if (cmfd_on)
-
-        **integer(4) cmfd % indices**
-
-            Indices for cmfd mesh (i,j,k,g)
-
-        **real(8) cmfd % k_cmfd(1:current_batch)**
-
-            CMFD eigenvalues
-
-        **real(8) cmfd % src(1:G,1:I,1:J,1:K)**
-
-            CMFD fission source
-
-        **real(8) cmfd % entropy(1:current_batch)**
-
-            CMFD estimate of Shannon entropy
-
-        **real(8) cmfd % balance(1:current_batch)**
-
-            RMS of the residual neutron balance equation on CMFD mesh
-
-        **real(8) cmfd % dom(1:current_batch)**
-
-            CMFD estimate of dominance ratio
-
-        **real(8) cmfd % scr_cmp(1:current_batch)**
-
-            RMS comparison of difference between OpenMC and CMFD fission source
-
-**integer(4) n_meshes**
-
-    Number of meshes in tallies.xml file
-
-*do i = 1, n_meshes*
-
-    **integer(4) meshes(i) % id**
-
-        Unique ID of mesh.
-
-    **integer(4) meshes(i) % type**
-
-        Type of mesh.
-
-    **integer(4) meshes(i) % n_dimension**
-
-        Number of dimensions for mesh (2 or 3).
-
-    **integer(4) meshes(i) % dimension(:)**
-
-        Number of mesh cells in each dimension.
-
-    **real(8) meshes(i) % lower_left(:)**
-
-        Coordinates of lower-left corner of mesh.
-
-    **real(8) meshes(i) % upper_right(:)**
-
-        Coordinates of upper-right corner of mesh.
-
-    **real(8) meshes(i) % width(:)**
-
-        Width of each mesh cell in each dimension.
-
-**integer(4) n_tallies**
-
-*do i = 1, n_tallies*
-
-    **integer(4) tallies(i) % id**
-
-        Unique ID of tally.
-
-    **integer(4) tallies(i) % n_realizations**
-
-        Number of realizations for the i-th tally.
-
-    **integer(4) size(tallies(i) % scores, 1)**
-
-        Total number of score bins for the i-th tally
-
-    **integer(4) size(tallies(i) % scores, 2)**
-
-        Total number of filter bins for the i-th tally
-
-    **integer(4) tallies(i) % n_filters**
-
-    *do j = 1, tallies(i) % n_filters*
-
-        **integer(4) tallies(i) % filter(j) % type**
-
-            Type of tally filter.
-
-        **integer(4) tallies(i) % filter(j) % n_bins**
-
-            Number of bins for filter.
-
-        **integer(4)/real(8) tallies(i) % filter(j) % bins(:)**
-
-            Value for each filter bin of this type.
-
-    **integer(4) tallies(i) % n_nuclide_bins**
-
-        Number of nuclide bins. If none are specified, this is just one.
-
-    *do j = 1, tallies(i) % n_nuclide_bins*
-
-        **integer(4) tallies(i) % nuclide_bins(j)**
-
-            Values of specified nuclide bins
-
-    **integer(4) tallies(i) % n_score_bins**
-
-        Number of scoring bins.
-
-    *do j = 1, tallies(i) % n_score_bins*
-
-        **integer(4) tallies(i) % score_bins(j)**
-
-            Values of specified scoring bins (e.g. SCORE_FLUX).
-
-    **integer(4) tallies(i) % n_score_bins**
-
-        Number of scoring bins without accounting for those added by
-        the scatter-pn command.
-
-    *do j = 1, tallies(i) % n_user_score_bins*
-
-        **character(8) tallies(i) % moment_order(j)**
-
-            Tallying moment order for Legendre and spherical
-            harmonic tally expansions (*e.g.*, 'P2', 'Y1,2', etc.).
-
-**integer(4) source_present**
-
-    Flag indicated if source bank is present in the file
-
-**integer(4) n_realizations**
-
-    Number of realizations for global tallies.
-
-**integer(4) N_GLOBAL_TALLIES**
-
-    Number of global tally scores
-
-*do i = 1, N_GLOBAL_TALLIES*
-
-    **real(8) global_tallies(i) % sum**
-
-        Accumulated sum for the i-th global tally
-
-    **real(8) global_tallies(i) % sum_sq**
-
-        Accumulated sum of squares for the i-th global tally
-
-**integer(4) tallies_on**
-
-    Flag indicated if tallies are present in the file.
-
-if (tallies_on > 0)
-
-    *do i = 1, n_tallies*
-
-        *do k = 1, size(tallies(i) % scores, 2)*
-
-            *do j = 1, size(tallies(i) % scores, 1)*
-
-                **real(8) tallies(i) % scores(j,k) % sum**
-            
-                    Accumulated sum for the j-th score and k-th filter of the
-                    i-th tally
-
-                **real(8) tallies(i) % scores(j,k) % sum_sq**
-
-                    Accumulated sum of squares for the j-th score and k-th
-                    filter of the i-th tally
-
-if (run_mode == MODE_EIGENVALUE and source_present)
-
-    *do i = 1, n_particles*
-
-        **real(8) source_bank(i) % wgt**
-
-            Weight of the i-th source particle
-
-        **real(8) source_bank(i) % xyz(1:3)**
-
-            Coordinates of the i-th source particle.
-
-        **real(8) source_bank(i) % uvw(1:3)**
-
-            Direction of the i-th source particle
-
-        **real(8) source_bank(i) % E**
-
-            Energy of the i-th source particle.
-

_sources/devguide/voxel.txt

@@ -1,52 +0,0 @@
-.. _devguide_voxel:
-
-=====================================
-Voxel Plot Binary File Specifications
-=====================================
-
-The current revision of the voxel plot binary file is 1.
-
-**integer(4) n_voxels_x**
-
-    Number of voxels in the x direction
-
-**integer(4) n_voxels_y**
-
-    Number of voxels in the y direction
-
-**integer(4) n_voxels_z**
-
-    Number of voxels in the z direction
-
-**real(8) width_voxel_x**
-
-    Width of voxels in the x direction
-
-**real(8) width_voxel_y**
-
-    Width of voxels in the y direction
-
-**real(8) width_voxel_z**
-
-    Width of voxels in the z direction
-
-**real(8) lower_left_x**
-
-    Lower left x point of the voxel grid
-
-**real(8) lower_left_y**
-
-    Lower left y point of the voxel grid
-
-**real(8) lower_left_z**
-
-    Lower left z point of the voxel grid
-
-*do x = 1, n_voxels_x*
-    *do y = 1, n_voxels_y*
-        *do z = 1, n_voxels_z*
-
-            **integer(4) id**
-
-                Cell or material id number at this voxel center. Set to -1 when
-                cell not_found.

_sources/methods/eigenvalue.txt

@@ -142,7 +142,7 @@ than unity. By ensuring that the expected number of fission sites in each mesh
 cell is constant, the collision density across all cells, and hence the variance
 of tallies, is more uniform than it would be otherwise.
 
-.. _Shannon entropy: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-06-3737_entropy.pdf
+.. _Shannon entropy: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-06-3737.pdf
 
 .. [Lieberoth] J. Lieberoth, "A Monte Carlo Technique to Solve the Static
    Eigenvalue Problem of the Boltzmann Transport Equation," *Nukleonik*, **11**,

_sources/methods/geometry.txt

@@ -10,7 +10,7 @@ Constructive Solid Geometry
 
 OpenMC uses a technique known as `constructive solid geometry`_ (CSG) to build
 arbitrarily complex three-dimensional models in Euclidean space. In a CSG model,
-every unique object is described as the union, intersection, or difference of
+every unique object is described as the union and/or intersection of
 *half-spaces* created by bounding `surfaces`_. Every surface divides all of
 space into exactly two half-spaces. We can mathematically define a surface as a
 collection of points that satisfy an equation of the form :math:`f(x,y,z) = 0`
@@ -54,13 +54,12 @@ dividing space into two half-spaces.
    Example of an ellipse and its associated half-spaces.
 
 References to half-spaces created by surfaces are used to define regions of
-space of uniform composition, known as *cells*. While some codes allow regions
-to be defined by intersections, unions, and differences or half-spaces, OpenMC
-is currently limited to cells defined only as intersections of
-half-spaces. Thus, the specification of the cell must include a list of
-half-space references whose intersection defines the region. The region is then
-assigned a material defined elsewhere. Figure :num:`fig-union` shows an
-example of a cell defined as the intersection of an ellipse and two planes.
+space of uniform composition, which are then assigned to *cells*. OpenMC allows
+regions to be defined using union, intersection, and complement operators. As in
+MCNP_, the intersection operator is implicit as doesn't need to be written in a
+region specification. A defined region is then associated with a material
+composition in a cell. Figure :num:`fig-union` shows an example of a cell region
+defined as the intersection of an ellipse and two planes.
 
 .. _fig-union:
 
@@ -117,6 +116,10 @@ to fully define the surface.
     | Cone parallel to the | z-cone     | :math:`(x-x_0)^2 + (y-y_0)^2 | :math:`x_0 \; y_0 \;    |
     | :math:`z`-axis       |            | = R^2(z-z_0)^2`              | z_0 \; R^2`             |
     +----------------------+------------+------------------------------+-------------------------+
+    | General quadric      | quadric    | :math:`Ax^2 + By^2 + Cz^2 +  | :math:`A \; B \; C \; D |
+    | surface              |            | Dxy + Eyz + Fxz + Gx + Hy +  | \; E \; F \; G \; H \;  |
+    |                      |            | Jz + K`                      | J \; K`                 |
+    +----------------------+------------+------------------------------+-------------------------+
 
 .. _universes:
 

_sources/methods/physics.txt

@@ -187,11 +187,11 @@ secondary photons from nuclear de-excitation are tracked in OpenMC.
 ------------------------
 
 These types of reactions are just treated as inelastic scattering and as such
-are subject to the same procedure as described in
-:ref:`inelastic-scatter`. Rather than tracking multiple secondary neutrons, the
-weight of the outgoing neutron is multiplied by the number of secondary
-neutrons, e.g. for :math:`(n,2n)`, only one outgoing neutron is tracked but its
-weight is doubled.
+are subject to the same procedure as described in :ref:`inelastic-scatter`. For
+reactions with integral multiplicity, e.g., :math:`(n,2n)`, an appropriate
+number of secondary neutrons are created. For reactions that have a multiplicity
+given as a function of the incoming neutron energy (which occasionally occurs
+for MT=5), the weight of the outgoing neutron is multiplied by the multiplcity.
 
 .. _fission:
 
@@ -1027,14 +1027,19 @@ probability distribution function can be found by integrating equation
 Let us call the normalization factor in the denominator of equation
 :eq:`target-pdf-1` :math:`C`.
 
-It is normally assumed that :math:`\sigma (v_r)` is constant over the range of
+
+Constant Cross Section Model
+----------------------------
+
+It is often assumed that :math:`\sigma (v_r)` is constant over the range of
 relative velocities of interest. This is a good assumption for almost all cases
 since the elastic scattering cross section varies slowly with velocity for light
 nuclei, and for heavy nuclei where large variations can occur due to resonance
 scattering, the moderating effect is rather small. Nonetheless, this assumption
 may cause incorrect answers in systems with low-lying resonances that can cause
 a significant amount of up-scatter that would be ignored by this assumption
-(e.g. U-238 in commercial light-water reactors). Nevertheless, with this
+(e.g. U-238 in commercial light-water reactors). We will revisit this assumption
+later in :ref:`energy_dependent_xs_model`. For now, continuing with the
 assumption, we write :math:`\sigma (v_r) = \sigma_s` which simplifies
 :eq:`target-pdf-1` to
 
@@ -1232,6 +1237,35 @@ If is not accepted, then we repeat the process and resample a target speed and
 cosine until a combination is found that satisfies equation
 :eq:`freegas-accept-2`.
 
+.. _energy_dependent_xs_model:
+
+Energy-Dependent Cross Section Model
+------------------------------------
+
+As was noted earlier, assuming that the elastic scattering cross section is
+constant in :eq:`reaction-rate` is not strictly correct, especially when
+low-lying resonances are present in the cross sections for heavy nuclides. To
+correctly account for energy dependence of the scattering cross section entails
+performing another rejection step. The most common method is to sample
+:math:`\mu` and :math:`v_T` as in the constant cross section approximation and
+then perform a rejection on the ratio of the 0 K elastic scattering cross
+section at the relative velocity to the maximum 0 K elastic scattering cross
+section over the range of velocities considered:
+
+.. math::
+    :label: dbrc
+
+    p_{dbrc} = \frac{\sigma_s(v_r)}{\sigma_{s,max}}
+
+where it should be noted that the maximum is taken over the range :math:`[v_n -
+4/\beta, 4_n + 4\beta]`. This method is known as Doppler broadening rejection
+correction (DBRC) and was first introduced by `Becker et al.`_. OpenMC has an
+implementation of DBRC as well as an accelerated sampling method that are
+described fully in `Walsh et al.`_
+
+.. _Becker et al.: http://dx.doi.org/10.1016/j.anucene.2008.12.001
+.. _Walsh et al.: http://dx.doi.org/10.1016/j.anucene.2014.01.017
+
 .. _sab_tables:
 
 ------------

_sources/publications.txt

@@ -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,

_sources/pythonapi/energy_groups.txt

@@ -0,0 +1,8 @@
+.. _pythonapi_energy_groups:
+
+=============
+Energy Groups
+=============
+
+.. automodule:: openmc.mgxs.groups
+    :members:

_sources/pythonapi/examples/mgxs-part-i-content.txt

@@ -0,0 +1,910 @@
+
+This IPython Notebook introduces the use of the ``openmc.mgxs`` module
+to calculate multi-group cross sections for an infinite homogeneous
+medium. In particular, this Notebook introduces the the following
+features:
+
+-  **General equations** for scalar-flux averaged multi-group cross
+   sections
+-  Creation of multi-group cross sections for an **infinite homogeneous
+   medium**
+-  Use of **tally arithmetic** to manipulate multi-group cross sections
+
+**Note:** This Notebook illustrates the use of
+`Pandas <http://pandas.pydata.org/>`__ ``DataFrames`` to containerize
+multi-group cross section data. We recommend using
+`Pandas <http://pandas.pydata.org/>`__ >v0.15.0 or later since OpenMC's
+Python API leverages the multi-indexing feature included in the most
+recent releases of `Pandas <http://pandas.pydata.org/>`__.
+
+Introduction to Multi-Group Cross Sections (MGXS)
+-------------------------------------------------
+
+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.
+
+.. code:: python
+
+    from IPython.display import Image
+    Image(filename='images/mgxs.png', width=350)
+
+
+
+
+.. image:: mgxs-part-i-content_files/mgxs-part-i-content_3_0.png
+
+
+
+A variety of tools employing different methodologies have been developed
+over the years to compute multi-group cross sections for certain
+applications, including NJOY (LANL), MC\ :math:`^2`-3 (ANL), and Serpent
+(VTT). The ``openmc.mgxs`` Python module is designed to leverage
+OpenMC's tally system to calculate multi-group cross sections with
+arbitrary energy discretizations for fine-mesh heterogeneous
+deterministic neutron transport applications.
+
+Before proceeding to illustrate how one may use the ``openmc.mgxs``
+module, it is worthwhile to define the general equations used to
+calculate multi-group cross sections. This is only intended as a brief
+overview of the methodology used by ``openmc.mgxs`` - we refer the
+interested reader to the large body of literature on the subject for a
+more comprehensive understanding of this complex topic.
+
+Introductory Notation
+~~~~~~~~~~~~~~~~~~~~~
+
+The continuous real-valued microscopic cross section may be denoted
+:math:`\sigma_{n,x}(\mathbf{r}, E)` for position vector
+:math:`\mathbf{r}`, energy :math:`E`, nuclide :math:`n` and interaction
+type :math:`x`. Similarly, the scalar neutron flux may be denoted by
+:math:`\Phi(\mathbf{r},E)` for position :math:`\mathbf{r}` and energy
+:math:`E`. **Note**: Although nuclear cross sections are dependent on
+the temperature :math:`T` of the interacting medium, the temperature
+variable is neglected here for brevity.
+
+Spatial and Energy Discretization
+~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
+
+The energy domain for critical systems such as thermal reactors spans
+more than 10 orders of magnitude of neutron energies from
+10\ :math:`^{-5}` - 10\ :math:`^7` eV. The multi-group approximation
+discretization divides this energy range into one or more energy groups.
+In particular, for :math:`G` total groups, we denote an energy group
+index :math:`g` such that :math:`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**.
+
+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 :math:`K` distinct spatial zones, we designate each
+spatial zone an index :math:`k` such that
+:math:`k \in \{1, 2, ..., K\}`. The volume of each spatial zone is
+denoted by :math:`V_{k}`. The integration over discrete spatial zones is
+commonly referred to as **spatial homogenization**.
+
+General Scalar-Flux Weighted MGXS
+~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
+
+The multi-group cross sections computed by ``openmc.mgxs`` are defined
+as a *scalar flux-weighted average* of the microscopic cross sections
+across each discrete energy group. This formulation is employed in order
+to preserve the reaction rates within each energy group and spatial
+zone. In particular, spatial homogenization and energy condensation are
+used to compute the general multi-group cross section
+:math:`\sigma_{n,x,k,g}` as follows:
+
+.. math:: \sigma_{n,x,k,g} = \frac{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\sigma_{n,x}(\mathbf{r},E')\Phi(\mathbf{r},E')}{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\Phi(\mathbf{r},E')}
+
+This scalar flux-weighted average microscopic cross section is computed
+by ``openmc.mgxs`` for most multi-group cross sections, including total,
+absorption, and fission reaction types. These double integrals are
+stochastically computed with OpenMC's tally system - in particular,
+`filters <https://mit-crpg.github.io/openmc/pythonapi/filter.html>`__ on
+the energy range and spatial zone (material, cell or universe) define
+the bounds of integration for both numerator and denominator.
+
+Multi-Group Scattering Matrices
+~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
+
+The general multi-group cross section :math:`\sigma_{n,x,k,g}` is a
+vector of :math:`G` values for each energy group :math:`g`. The equation
+presented above only discretizes the energy of the incoming neutron and
+neglects the outgoing energy of the neutron (if any). Hence, this
+formulation must be extended to account for the outgoing energy of
+neutrons in the discretized scattering matrix cross section used by
+deterministic neutron transport codes.
+
+We denote the incoming and outgoing neutron energy groups as :math:`g`
+and :math:`g'` for the microscopic scattering matrix cross section
+:math:`\sigma_{n,s}(\mathbf{r},E)`. As before, spatial homogenization
+and energy condensation are used to find the multi-group scattering
+matrix cross section :math:`\sigma_{n,s,k,g \to g'}` as follows:
+
+.. math:: \sigma_{n,s,k,g\rightarrow g'} = \frac{\int_{E_{g'}}^{E_{g'-1}}\mathrm{d}E''\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\sigma_{n,s}(\mathbf{r},E'\rightarrow E'')\Phi(\mathbf{r},E')}{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\Phi(\mathbf{r},E')}
+
+This scalar flux-weighted multi-group microscopic scattering matrix is
+computed using OpenMC tallies with both energy in and energy out
+filters.
+
+Multi-Group Fission Spectrum
+~~~~~~~~~~~~~~~~~~~~~~~~~~~~
+
+The energy spectrum of neutrons emitted from fission is denoted by
+:math:`\chi_{n}(\mathbf{r},E' \rightarrow E'')` for incoming and
+outgoing energies :math:`E'` and :math:`E''`, respectively. Unlike the
+multi-group cross sections :math:`\sigma_{n,x,k,g}` considered up to
+this point, the fission spectrum is a probability distribution and must
+sum to unity. The outgoing energy is typically much less dependent on
+the incoming energy for fission than for scattering interactions. As a
+result, it is common practice to integrate over the incoming neutron
+energy when computing the multi-group fission spectrum. The fission
+spectrum may be simplified as :math:`\chi_{n}(\mathbf{r},E)` with
+outgoing energy :math:`E`.
+
+Unlike the multi-group cross sections defined up to this point, the
+multi-group fission spectrum is weighted by the fission production rate
+rather than the scalar flux. This formulation is intended to preserve
+the total fission production rate in the multi-group deterministic
+calculation. In order to mathematically define the multi-group fission
+spectrum, we denote the microscopic fission cross section as
+:math:`\sigma_{n,f}(\mathbf{r},E)` and the average number of neutrons
+emitted from fission interactions with nuclide :math:`n` as
+:math:`\nu_{n}(\mathbf{r},E)`. The multi-group fission spectrum
+:math:`\chi_{n,k,g}` is then the probability of fission neutrons emitted
+into energy group :math:`g`.
+
+Similar to before, spatial homogenization and energy condensation are
+used to find the multi-group fission spectrum :math:`\chi_{n,k,g}` as
+follows:
+
+.. math:: \chi_{n,k,g'} = \frac{\int_{E_{g'}}^{E_{g'-1}}\mathrm{d}E''\int_{0}^{\infty}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\chi_{n}(\mathbf{r},E'\rightarrow E'')\nu_{n}(\mathbf{r},E')\sigma_{n,f}(\mathbf{r},E')\Phi(\mathbf{r},E')}{\int_{0}^{\infty}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\nu_{n}(\mathbf{r},E')\sigma_{n,f}(\mathbf{r},E')\Phi(\mathbf{r},E')}
+
+The fission production-weighted multi-group fission spectrum is computed
+using OpenMC tallies with both energy in and energy out filters.
+
+This concludes our brief overview on the methodology to compute
+multi-group cross sections. The following sections detail more
+concretely how users may employ the ``openmc.mgxs`` module to power
+simulation workflows requiring multi-group cross sections for downstream
+deterministic calculations.
+
+Generate Input Files
+--------------------
+
+.. code:: python
+
+    import numpy as np
+    import matplotlib.pyplot as plt
+    
+    import openmc
+    import openmc.mgxs as mgxs
+    
+    %matplotlib inline
+
+First we need to define materials that will be used in the problem.
+Before defining a material, we must create nuclides that are used in the
+material.
+
+.. code:: python
+
+    # Instantiate some Nuclides
+    h1 = openmc.Nuclide('H-1')
+    o16 = openmc.Nuclide('O-16')
+    u235 = openmc.Nuclide('U-235')
+    u238 = openmc.Nuclide('U-238')
+    zr90 = openmc.Nuclide('Zr-90')
+
+With the nuclides we defined, we will now create a material for the
+homogeneous medium.
+
+.. code:: python
+
+    # Instantiate a Material and register the Nuclides
+    inf_medium = openmc.Material(name='moderator')
+    inf_medium.set_density('g/cc', 5.)
+    inf_medium.add_nuclide(h1,  0.028999667)
+    inf_medium.add_nuclide(o16, 0.01450188)
+    inf_medium.add_nuclide(u235, 0.000114142)
+    inf_medium.add_nuclide(u238, 0.006886019)
+    inf_medium.add_nuclide(zr90, 0.002116053)
+
+With our material, we can now create a ``MaterialsFile`` object that can
+be exported to an actual XML file.
+
+.. code:: python
+
+    # Instantiate a MaterialsFile, register all Materials, and export to XML
+    materials_file = openmc.MaterialsFile()
+    materials_file.default_xs = '71c'
+    materials_file.add_material(inf_medium)
+    materials_file.export_to_xml()
+
+Now let's move on to the geometry. This problem will be a simple square
+cell with reflective boundary conditions to simulate an infinite
+homogeneous medium. The first step is to create the outer bounding
+surfaces of the problem.
+
+.. code:: python
+
+    # Instantiate boundary Planes
+    min_x = openmc.XPlane(boundary_type='reflective', x0=-0.63)
+    max_x = openmc.XPlane(boundary_type='reflective', x0=0.63)
+    min_y = openmc.YPlane(boundary_type='reflective', y0=-0.63)
+    max_y = openmc.YPlane(boundary_type='reflective', y0=0.63)
+
+With the surfaces defined, we can now create a cell that is defined by
+intersections of half-spaces created by the surfaces.
+
+.. code:: python
+
+    # Instantiate a Cell
+    cell = openmc.Cell(cell_id=1, name='cell')
+    
+    # Register bounding Surfaces with the Cell
+    cell.region = +min_x & -max_x & +min_y & -max_y
+    
+    # Fill the Cell with the Material
+    cell.fill = inf_medium
+
+OpenMC requires that there is a "root" universe. Let us create a root
+universe and add our square cell to it.
+
+.. code:: python
+
+    # Instantiate Universe
+    root_universe = openmc.Universe(universe_id=0, name='root universe')
+    root_universe.add_cell(cell)
+
+We now must create a geometry that is assigned a root universe, put the
+geometry into a ``GeometryFile`` object, and export it to XML.
+
+.. code:: python
+
+    # Create Geometry and set root Universe
+    openmc_geometry = openmc.Geometry()
+    openmc_geometry.root_universe = root_universe
+    
+    # Instantiate a GeometryFile
+    geometry_file = openmc.GeometryFile()
+    geometry_file.geometry = openmc_geometry
+    
+    # Export to "geometry.xml"
+    geometry_file.export_to_xml()
+
+Next, we must define simulation parameters. In this case, we will use 10
+inactive batches and 40 active batches each with 2500 particles.
+
+.. code:: python
+
+    # OpenMC simulation parameters
+    batches = 50
+    inactive = 10
+    particles = 2500
+    
+    # Instantiate a SettingsFile
+    settings_file = openmc.SettingsFile()
+    settings_file.batches = batches
+    settings_file.inactive = inactive
+    settings_file.particles = particles
+    settings_file.output = {'tallies': True, 'summary': True}
+    bounds = [-0.63, -0.63, -0.63, 0.63, 0.63, 0.63]
+    settings_file.set_source_space('fission', bounds)
+    
+    # Export to "settings.xml"
+    settings_file.export_to_xml()
+
+Now we are ready to generate multi-group cross sections! First, let's
+define a 2-group structure using the built-in ``EnergyGroups`` class.
+
+.. code:: python
+
+    # Instantiate a 2-group EnergyGroups object
+    groups = mgxs.EnergyGroups()
+    groups.group_edges = np.array([0., 0.625e-6, 20.])
+
+We can now use the ``EnergyGroups`` object, along with our previously
+created materials and geometry, to instantiate some ``MGXS`` objects
+from the ``openmc.mgxs`` module. In particular, the following are
+subclasses of the generic and abstract ``MGXS`` class:
+
+-  ``TotalXS``
+-  ``TransportXS``
+-  ``AbsorptionXS``
+-  ``CaptureXS``
+-  ``FissionXS``
+-  ``NuFissionXS``
+-  ``ScatterXS``
+-  ``NuScatterXS``
+-  ``ScatterMatrixXS``
+-  ``NuScatterMatrixXS``
+-  ``Chi``
+
+These classes provide us with an interface to generate the tally inputs
+as well as perform post-processing of OpenMC's tally data to compute the
+respective multi-group cross sections. In this case, let's create the
+multi-group total, absorption and scattering cross sections with our
+2-group structure.
+
+.. code:: python
+
+    # Instantiate a few different sections
+    total = mgxs.TotalXS(domain=cell, domain_type='cell', groups=groups)
+    absorption = mgxs.AbsorptionXS(domain=cell, domain_type='cell', groups=groups)
+    scattering = mgxs.ScatterXS(domain=cell, domain_type='cell', groups=groups)
+
+Each multi-group cross section object stores its tallies in a Python
+dictionary called ``tallies``. We can inspect the tallies in the
+dictionary for our ``Absorption`` object as follows.
+
+.. code:: python
+
+    absorption.tallies
+
+
+
+
+.. parsed-literal::
+
+    OrderedDict([('flux', Tally
+    	ID             =	10000
+    	Name           =	
+    	Filters        =	
+                    		cell	[1]
+                    		energy	[  0.00000000e+00   6.25000000e-07   2.00000000e+01]
+    	Nuclides       =	total 
+    	Scores         =	['flux']
+    	Estimator      =	tracklength
+    ), ('absorption', Tally
+    	ID             =	10001
+    	Name           =	
+    	Filters        =	
+                    		cell	[1]
+                    		energy	[  0.00000000e+00   6.25000000e-07   2.00000000e+01]
+    	Nuclides       =	total 
+    	Scores         =	['absorption']
+    	Estimator      =	tracklength
+    )])
+
+
+
+The ``Absorption`` object includes tracklength tallies for the
+'absorption' and 'flux' scores in the 2-group structure in cell 1. Now
+that each ``MGXS`` object contains the tallies that it needs, we must
+add these tallies to a ``TalliesFile`` object to generate the
+"tallies.xml" input file for OpenMC.
+
+.. code:: python
+
+    # Instantiate an empty TalliesFile
+    tallies_file = openmc.TalliesFile()
+    
+    # Add total tallies to the tallies file
+    for tally in total.tallies.values():
+        tallies_file.add_tally(tally)
+    
+    # Add absorption tallies to the tallies file
+    for tally in absorption.tallies.values():
+        tallies_file.add_tally(tally)
+    
+    # Add scattering tallies to the tallies file
+    for tally in scattering.tallies.values():
+        tallies_file.add_tally(tally)
+                    
+    # Export to "tallies.xml"
+    tallies_file.export_to_xml()
+
+Now we a have a complete set of inputs, so we can go ahead and run our
+simulation.
+
+.. code:: python
+
+    # Run OpenMC
+    executor = openmc.Executor()
+    executor.run_simulation()
+
+
+.. parsed-literal::
+
+    
+           .d88888b.                             888b     d888  .d8888b.
+          d88P" "Y88b                            8888b   d8888 d88P  Y88b
+          888     888                            88888b.d88888 888    888
+          888     888 88888b.   .d88b.  88888b.  888Y88888P888 888       
+          888     888 888 "88b d8P  Y8b 888 "88b 888 Y888P 888 888       
+          888     888 888  888 88888888 888  888 888  Y8P  888 888    888
+          Y88b. .d88P 888 d88P Y8b.     888  888 888   "   888 Y88b  d88P
+           "Y88888P"  88888P"   "Y8888  888  888 888       888  "Y8888P"
+    __________________888______________________________________________________
+                      888
+                      888
+    
+          Copyright:      2011-2015 Massachusetts Institute of Technology
+          License:        http://mit-crpg.github.io/openmc/license.html
+          Version:        0.7.0
+          Git SHA1:       c4b14a5ef87f004528d35cbf33fef3ed15a386ca
+          Date/Time:      2015-12-02 09:11:05
+          MPI Processes:  1
+    
+     ===========================================================================
+     ========================>     INITIALIZATION     <=========================
+     ===========================================================================
+    
+     Reading settings XML file...
+     Reading cross sections XML file...
+     Reading geometry XML file...
+     Reading materials XML file...
+     Reading tallies XML file...
+     Building neighboring cells lists for each surface...
+     Loading ACE cross section table: 1001.71c
+     Loading ACE cross section table: 8016.71c
+     Loading ACE cross section table: 92235.71c
+     Loading ACE cross section table: 92238.71c
+     Loading ACE cross section table: 40090.71c
+     Maximum neutron transport energy: 20.0000 MeV for 1001.71c
+     Initializing source particles...
+    
+     ===========================================================================
+     ====================>     K EIGENVALUE SIMULATION     <====================
+     ===========================================================================
+    
+      Bat./Gen.      k            Average k         
+      =========   ========   ====================   
+            1/1    1.19804                       
+            2/1    1.12945                       
+            3/1    1.15573                       
+            4/1    1.13929                       
+            5/1    1.16300                       
+            6/1    1.22117                       
+            7/1    1.19012                       
+            8/1    1.11299                       
+            9/1    1.16066                       
+           10/1    1.12566                       
+           11/1    1.20854                       
+           12/1    1.14691    1.17773 +/- 0.03082
+           13/1    1.17204    1.17583 +/- 0.01789
+           14/1    1.14148    1.16724 +/- 0.01529
+           15/1    1.17272    1.16834 +/- 0.01189
+           16/1    1.18575    1.17124 +/- 0.01014
+           17/1    1.20498    1.17606 +/- 0.00983
+           18/1    1.14754    1.17249 +/- 0.00923
+           19/1    1.18141    1.17348 +/- 0.00820
+           20/1    1.15074    1.17121 +/- 0.00768
+           21/1    1.15914    1.17011 +/- 0.00703
+           22/1    1.14586    1.16809 +/- 0.00673
+           23/1    1.18999    1.16978 +/- 0.00642
+           24/1    1.15101    1.16844 +/- 0.00609
+           25/1    1.13791    1.16640 +/- 0.00602
+           26/1    1.19791    1.16837 +/- 0.00597
+           27/1    1.19818    1.17012 +/- 0.00587
+           28/1    1.14160    1.16854 +/- 0.00576
+           29/1    1.11487    1.16571 +/- 0.00614
+           30/1    1.17538    1.16620 +/- 0.00584
+           31/1    1.20210    1.16791 +/- 0.00581
+           32/1    1.20078    1.16940 +/- 0.00574
+           33/1    1.14624    1.16839 +/- 0.00558
+           34/1    1.14618    1.16747 +/- 0.00542
+           35/1    1.16866    1.16752 +/- 0.00520
+           36/1    1.18565    1.16821 +/- 0.00504
+           37/1    1.16824    1.16821 +/- 0.00485
+           38/1    1.18299    1.16874 +/- 0.00471
+           39/1    1.21418    1.17031 +/- 0.00480
+           40/1    1.11167    1.16835 +/- 0.00504
+           41/1    1.11545    1.16665 +/- 0.00516
+           42/1    1.11114    1.16491 +/- 0.00529
+           43/1    1.14227    1.16423 +/- 0.00517
+           44/1    1.14104    1.16355 +/- 0.00506
+           45/1    1.16756    1.16366 +/- 0.00492
+           46/1    1.13065    1.16274 +/- 0.00487
+           47/1    1.11251    1.16139 +/- 0.00492
+           48/1    1.14731    1.16101 +/- 0.00481
+           49/1    1.16691    1.16117 +/- 0.00469
+           50/1    1.19679    1.16206 +/- 0.00465
+     Creating state point statepoint.50.h5...
+    
+     ===========================================================================
+     ======================>     SIMULATION FINISHED     <======================
+     ===========================================================================
+    
+    
+     =======================>     TIMING STATISTICS     <=======================
+    
+     Total time for initialization     =  4.1700E-01 seconds
+       Reading cross sections          =  8.9000E-02 seconds
+     Total time in simulation          =  1.4728E+01 seconds
+       Time in transport only          =  1.4712E+01 seconds
+       Time in inactive batches        =  1.7890E+00 seconds
+       Time in active batches          =  1.2939E+01 seconds
+       Time synchronizing fission bank =  5.0000E-03 seconds
+         Sampling source sites         =  3.0000E-03 seconds
+         SEND/RECV source sites        =  2.0000E-03 seconds
+       Time accumulating tallies       =  1.0000E-03 seconds
+     Total time for finalization       =  1.0000E-03 seconds
+     Total time elapsed                =  1.5155E+01 seconds
+     Calculation Rate (inactive)       =  13974.3 neutrons/second
+     Calculation Rate (active)         =  7728.57 neutrons/second
+    
+     ============================>     RESULTS     <============================
+    
+     k-effective (Collision)     =  1.16131 +/-  0.00453
+     k-effective (Track-length)  =  1.16206 +/-  0.00465
+     k-effective (Absorption)    =  1.16096 +/-  0.00364
+     Combined k-effective        =  1.16120 +/-  0.00325
+     Leakage Fraction            =  0.00000 +/-  0.00000
+    
+
+
+
+
+.. parsed-literal::
+
+    0
+
+
+
+Tally Data Processing
+---------------------
+
+Our simulation ran successfully and created statepoint and summary
+output files. We begin our analysis by instantiating a ``StatePoint``
+object.
+
+.. code:: python
+
+    # Load the last statepoint file
+    sp = openmc.StatePoint('statepoint.50.h5')
+
+In addition to the statepoint file, our simulation also created a
+summary file which encapsulates information about the materials and
+geometry. This is necessary for the ``openmc.mgxs`` module to properly
+process the tally data. We first create a ``Summary`` object and link it
+with the statepoint.
+
+.. code:: python
+
+    # Load the summary file and link it with the statepoint
+    su = openmc.Summary('summary.h5')
+    sp.link_with_summary(su)
+
+The statepoint is now ready to be analyzed by our multi-group cross
+sections. We simply have to load the tallies from the ``StatePoint``
+into each object as follows and our ``MGXS`` objects will compute the
+cross sections for us under-the-hood.
+
+.. code:: python
+
+    # Load the tallies from the statepoint into each MGXS object
+    total.load_from_statepoint(sp)
+    absorption.load_from_statepoint(sp)
+    scattering.load_from_statepoint(sp)
+
+Voila! Our multi-group cross sections are now ready to rock 'n roll!
+
+Extracting and Storing MGXS Data
+--------------------------------
+
+Let's first inspect our total cross section by printing it to the
+screen.
+
+.. code:: python
+
+    total.print_xs()
+
+
+.. parsed-literal::
+
+    Multi-Group XS
+    	Reaction Type  =	total
+    	Domain Type    =	cell
+    	Domain ID      =	1
+    	Cross Sections [cm^-1]:
+                Group 1 [6.25e-07   - 20.0      MeV]:	6.81e-01 +/- 1.88e-01%
+                Group 2 [0.0        - 6.25e-07  MeV]:	1.40e+00 +/- 5.91e-01%
+    
+    
+    
+
+
+Since the ``openmc.mgxs`` module uses `tally
+arithmetic <https://mit-crpg.github.io/openmc/pythonapi/examples/tally-arithmetic.html>`__
+under-the-hood, the cross section is stored as a "derived" ``Tally``
+object. This means that it can be queried and manipulated using all of
+the same methods supported for the ``Tally`` class in the OpenMC Python
+API. For example, we can construct a
+`Pandas <http://pandas.pydata.org/>`__ ``DataFrame`` of the multi-group
+cross section data.
+
+.. code:: python
+
+    df = scattering.get_pandas_dataframe()
+    df.head(10)
+
+
+
+
+.. raw:: html
+
+    <div>
+    <table border="1" class="dataframe">
+      <thead>
+        <tr style="text-align: right;">
+          <th></th>
+          <th>cell</th>
+          <th>group in</th>
+          <th>nuclide</th>
+          <th>mean</th>
+          <th>std. dev.</th>
+        </tr>
+      </thead>
+      <tbody>
+        <tr>
+          <th>1</th>
+          <td>1</td>
+          <td>1</td>
+          <td>total</td>
+          <td>0.668323</td>
+          <td>0.001264</td>
+        </tr>
+        <tr>
+          <th>0</th>
+          <td>1</td>
+          <td>2</td>
+          <td>total</td>
+          <td>1.293258</td>
+          <td>0.007624</td>
+        </tr>
+      </tbody>
+    </table>
+    </div>
+
+
+
+Each multi-group cross section object can be easily exported to a
+variety of file formats, including CSV, Excel, and LaTeX for storage or
+data processing.
+
+.. code:: python
+
+    absorption.export_xs_data(filename='absorption-xs', format='excel')
+
+The following code snippet shows how to export all three ``MGXS`` to the
+same HDF5 binary data store.
+
+.. code:: python
+
+    total.build_hdf5_store(filename='mgxs', append=True)
+    absorption.build_hdf5_store(filename='mgxs', append=True)
+    scattering.build_hdf5_store(filename='mgxs', append=True)
+
+Comparing MGXS with Tally Arithmetic
+------------------------------------
+
+Finally, we illustrate how one can leverage OpenMC's `tally
+arithmetic <https://mit-crpg.github.io/openmc/pythonapi/examples/tally-arithmetic.html>`__
+data processing feature with ``MGXS`` objects. The ``openmc.mgxs``
+module uses tally arithmetic to compute multi-group cross sections with
+automated uncertainty propagation. Each ``MGXS`` object includes an
+``xs_tally`` attribute which is a "derived" ``Tally`` based on the
+tallies needed to compute the cross section type of interest. These
+derived tallies can be used in subsequent tally arithmetic operations.
+For example, we can use tally artithmetic to confirm that the
+``TotalXS`` is equal to the sum of the ``AbsorptionXS`` and
+``ScatterXS`` objects.
+
+.. code:: python
+
+    # Use tally arithmetic to compute the difference between the total, absorption and scattering
+    difference = total.xs_tally - absorption.xs_tally - scattering.xs_tally
+    
+    # The difference is a derived tally which can generate Pandas DataFrames for inspection
+    difference.get_pandas_dataframe()
+
+
+
+
+.. raw:: html
+
+    <div>
+    <table border="1" class="dataframe">
+      <thead>
+        <tr style="text-align: right;">
+          <th></th>
+          <th>cell</th>
+          <th>energy [MeV]</th>
+          <th>nuclide</th>
+          <th>score</th>
+          <th>mean</th>
+          <th>std. dev.</th>
+        </tr>
+      </thead>
+      <tbody>
+        <tr>
+          <th>0</th>
+          <td>1</td>
+          <td>(0.0e+00 - 6.3e-07)</td>
+          <td>total</td>
+          <td>(((total / flux) - (absorption / flux)) - (sca...</td>
+          <td>4.884981e-15</td>
+          <td>0.011274</td>
+        </tr>
+        <tr>
+          <th>1</th>
+          <td>1</td>
+          <td>(6.3e-07 - 2.0e+01)</td>
+          <td>total</td>
+          <td>(((total / flux) - (absorption / flux)) - (sca...</td>
+          <td>1.221245e-15</td>
+          <td>0.001802</td>
+        </tr>
+      </tbody>
+    </table>
+    </div>
+
+
+
+Similarly, we can use tally arithmetic to compute the ratio of
+``AbsorptionXS`` and ``ScatterXS`` to the ``TotalXS``.
+
+.. code:: python
+
+    # Use tally arithmetic to compute the absorption-to-total MGXS ratio
+    absorption_to_total = absorption.xs_tally / total.xs_tally
+    
+    # The absorption-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection
+    absorption_to_total.get_pandas_dataframe()
+
+
+
+
+.. raw:: html
+
+    <div>
+    <table border="1" class="dataframe">
+      <thead>
+        <tr style="text-align: right;">
+          <th></th>
+          <th>cell</th>
+          <th>energy [MeV]</th>
+          <th>nuclide</th>
+          <th>score</th>
+          <th>mean</th>
+          <th>std. dev.</th>
+        </tr>
+      </thead>
+      <tbody>
+        <tr>
+          <th>0</th>
+          <td>1</td>
+          <td>(0.0e+00 - 6.3e-07)</td>
+          <td>total</td>
+          <td>((absorption / flux) / (total / flux))</td>
+          <td>0.076219</td>
+          <td>0.000651</td>
+        </tr>
+        <tr>
+          <th>1</th>
+          <td>1</td>
+          <td>(6.3e-07 - 2.0e+01)</td>
+          <td>total</td>
+          <td>((absorption / flux) / (total / flux))</td>
+          <td>0.019319</td>
+          <td>0.000086</td>
+        </tr>
+      </tbody>
+    </table>
+    </div>
+
+
+
+.. code:: python
+
+    # Use tally arithmetic to compute the scattering-to-total MGXS ratio
+    scattering_to_total = scattering.xs_tally / total.xs_tally
+    
+    # The scattering-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection
+    scattering_to_total.get_pandas_dataframe()
+
+
+
+
+.. raw:: html
+
+    <div>
+    <table border="1" class="dataframe">
+      <thead>
+        <tr style="text-align: right;">
+          <th></th>
+          <th>cell</th>
+          <th>energy [MeV]</th>
+          <th>nuclide</th>
+          <th>score</th>
+          <th>mean</th>
+          <th>std. dev.</th>
+        </tr>
+      </thead>
+      <tbody>
+        <tr>
+          <th>0</th>
+          <td>1</td>
+          <td>(0.0e+00 - 6.3e-07)</td>
+          <td>total</td>
+          <td>((scatter / flux) / (total / flux))</td>
+          <td>0.923781</td>
+          <td>0.007714</td>
+        </tr>
+        <tr>
+          <th>1</th>
+          <td>1</td>
+          <td>(6.3e-07 - 2.0e+01)</td>
+          <td>total</td>
+          <td>((scatter / flux) / (total / flux))</td>
+          <td>0.980681</td>
+          <td>0.002617</td>
+        </tr>
+      </tbody>
+    </table>
+    </div>
+
+
+
+Lastly, we sum the derived scatter-to-total and absorption-to-total
+ratios to confirm that they sum to unity.
+
+.. code:: python
+
+    # Use tally arithmetic to ensure that the absorption- and scattering-to-total MGXS ratios sum to unity
+    sum_ratio = absorption_to_total + scattering_to_total
+    
+    # The scattering-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection
+    sum_ratio.get_pandas_dataframe()
+
+
+
+
+.. raw:: html
+
+    <div>
+    <table border="1" class="dataframe">
+      <thead>
+        <tr style="text-align: right;">
+          <th></th>
+          <th>cell</th>
+          <th>energy [MeV]</th>
+          <th>nuclide</th>
+          <th>score</th>
+          <th>mean</th>
+          <th>std. dev.</th>
+        </tr>
+      </thead>
+      <tbody>
+        <tr>
+          <th>0</th>
+          <td>1</td>
+          <td>(0.0e+00 - 6.3e-07)</td>
+          <td>total</td>
+          <td>(((absorption / flux) / (total / flux)) + ((sc...</td>
+          <td>1</td>
+          <td>0.007741</td>
+        </tr>
+        <tr>
+          <th>1</th>
+          <td>1</td>
+          <td>(6.3e-07 - 2.0e+01)</td>
+          <td>total</td>
+          <td>(((absorption / flux) / (total / flux)) + ((sc...</td>
+          <td>1</td>
+          <td>0.002619</td>
+        </tr>
+      </tbody>
+    </table>
+    </div>
+
+

_sources/pythonapi/examples/mgxs-part-i.txt

@@ -0,0 +1,13 @@
+.. _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.

_sources/pythonapi/examples/mgxs-part-ii.txt

@@ -0,0 +1,13 @@
+.. _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.

_sources/pythonapi/examples/mgxs-part-iii.txt

@@ -0,0 +1,13 @@
+.. _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.

_sources/pythonapi/examples/post-processing-content.txt

@@ -0,0 +1,767 @@
+
+This notebook demonstrates some basic post-processing tasks that can be
+performed with the Python API, such as plotting a 2D mesh tally and
+plotting neutron source sites from an eigenvalue calculation. The
+problem we will use is a simple reflected pin-cell.
+
+.. code:: python
+
+    from IPython.display import Image
+    import numpy as np
+    import matplotlib.pyplot as plt
+    
+    import openmc
+    from openmc.statepoint import StatePoint
+    
+    %matplotlib inline
+
+Generate Input Files
+--------------------
+
+First we need to define materials that will be used in the problem.
+Before defining a material, we must create nuclides that are used in the
+material.
+
+.. code:: python
+
+    # Instantiate some Nuclides
+    h1 = openmc.Nuclide('H-1')
+    b10 = openmc.Nuclide('B-10')
+    o16 = openmc.Nuclide('O-16')
+    u235 = openmc.Nuclide('U-235')
+    u238 = openmc.Nuclide('U-238')
+    zr90 = openmc.Nuclide('Zr-90')
+
+With the nuclides we defined, we will now create three materials for the
+fuel, water, and cladding of the fuel pin.
+
+.. code:: python
+
+    # 1.6 enriched fuel
+    fuel = openmc.Material(name='1.6% Fuel')
+    fuel.set_density('g/cm3', 10.31341)
+    fuel.add_nuclide(u235, 3.7503e-4)
+    fuel.add_nuclide(u238, 2.2625e-2)
+    fuel.add_nuclide(o16, 4.6007e-2)
+    
+    # borated water
+    water = openmc.Material(name='Borated Water')
+    water.set_density('g/cm3', 0.740582)
+    water.add_nuclide(h1, 4.9457e-2)
+    water.add_nuclide(o16, 2.4732e-2)
+    water.add_nuclide(b10, 8.0042e-6)
+    
+    # zircaloy
+    zircaloy = openmc.Material(name='Zircaloy')
+    zircaloy.set_density('g/cm3', 6.55)
+    zircaloy.add_nuclide(zr90, 7.2758e-3)
+
+With our three materials, we can now create a materials file object that
+can be exported to an actual XML file.
+
+.. code:: python
+
+    # Instantiate a MaterialsFile, add Materials
+    materials_file = openmc.MaterialsFile()
+    materials_file.add_material(fuel)
+    materials_file.add_material(water)
+    materials_file.add_material(zircaloy)
+    materials_file.default_xs = '71c'
+    
+    # Export to "materials.xml"
+    materials_file.export_to_xml()
+
+Now let's move on to the geometry. Our problem will have three regions
+for the fuel, the clad, and the surrounding coolant. The first step is
+to create the bounding surfaces -- in this case two cylinders and six
+reflective planes.
+
+.. code:: python
+
+    # Create cylinders for the fuel and clad
+    fuel_outer_radius = openmc.ZCylinder(x0=0.0, y0=0.0, R=0.39218)
+    clad_outer_radius = openmc.ZCylinder(x0=0.0, y0=0.0, R=0.45720)
+    
+    # Create boundary planes to surround the geometry
+    # Use both reflective and vacuum boundaries to make life interesting
+    min_x = openmc.XPlane(x0=-0.63, boundary_type='reflective')
+    max_x = openmc.XPlane(x0=+0.63, boundary_type='reflective')
+    min_y = openmc.YPlane(y0=-0.63, boundary_type='reflective')
+    max_y = openmc.YPlane(y0=+0.63, boundary_type='reflective')
+    min_z = openmc.ZPlane(z0=-0.63, boundary_type='reflective')
+    max_z = openmc.ZPlane(z0=+0.63, boundary_type='reflective')
+
+With the surfaces defined, we can now create cells that are defined by
+intersections of half-spaces created by the surfaces.
+
+.. code:: python
+
+    # Create a Universe to encapsulate a fuel pin
+    pin_cell_universe = openmc.Universe(name='1.6% Fuel Pin')
+    
+    # Create fuel Cell
+    fuel_cell = openmc.Cell(name='1.6% Fuel')
+    fuel_cell.fill = fuel
+    fuel_cell.region = -fuel_outer_radius
+    pin_cell_universe.add_cell(fuel_cell)
+    
+    # Create a clad Cell
+    clad_cell = openmc.Cell(name='1.6% Clad')
+    clad_cell.fill = zircaloy
+    clad_cell.region = +fuel_outer_radius & -clad_outer_radius
+    pin_cell_universe.add_cell(clad_cell)
+    
+    # Create a moderator Cell
+    moderator_cell = openmc.Cell(name='1.6% Moderator')
+    moderator_cell.fill = water
+    moderator_cell.region = +clad_outer_radius
+    pin_cell_universe.add_cell(moderator_cell)
+
+OpenMC requires that there is a "root" universe. Let us create a root
+cell that is filled by the pin cell universe and then assign it to the
+root universe.
+
+.. code:: python
+
+    # Create root Cell
+    root_cell = openmc.Cell(name='root cell')
+    root_cell.fill = pin_cell_universe
+    
+    # Add boundary planes
+    root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
+    
+    # Create root Universe
+    root_universe = openmc.Universe(universe_id=0, name='root universe')
+    root_universe.add_cell(root_cell)
+
+We now must create a geometry that is assigned a root universe, put the
+geometry into a geometry file, and export it to XML.
+
+.. code:: python
+
+    # Create Geometry and set root Universe
+    geometry = openmc.Geometry()
+    geometry.root_universe = root_universe
+
+.. code:: python
+
+    # Instantiate a GeometryFile
+    geometry_file = openmc.GeometryFile()
+    geometry_file.geometry = geometry
+    
+    # Export to "geometry.xml"
+    geometry_file.export_to_xml()
+
+With the geometry and materials finished, we now just need to define
+simulation parameters. In this case, we will use 10 inactive batches and
+90 active batches each with 5000 particles.
+
+.. code:: python
+
+    # OpenMC simulation parameters
+    batches = 100
+    inactive = 10
+    particles = 5000
+    
+    # Instantiate a SettingsFile
+    settings_file = openmc.SettingsFile()
+    settings_file.batches = batches
+    settings_file.inactive = inactive
+    settings_file.particles = particles
+    source_bounds = [-0.63, -0.63, -0.63, 0.63, 0.63, 0.63]
+    settings_file.set_source_space('box', source_bounds)
+    
+    # Export to "settings.xml"
+    settings_file.export_to_xml()
+
+Let us also create a plot file that we can use to verify that our pin
+cell geometry was created successfully.
+
+.. code:: python
+
+    # Instantiate a Plot
+    plot = openmc.Plot(plot_id=1)
+    plot.filename = 'materials-xy'
+    plot.origin = [0, 0, 0]
+    plot.width = [1.26, 1.26]
+    plot.pixels = [250, 250]
+    plot.color = 'mat'
+    
+    # Instantiate a PlotsFile, add Plot, and export to "plots.xml"
+    plot_file = openmc.PlotsFile()
+    plot_file.add_plot(plot)
+    plot_file.export_to_xml()
+
+With the plots.xml file, we can now generate and view the plot. OpenMC
+outputs plots in .ppm format, which can be converted into a compressed
+format like .png with the convert utility.
+
+.. code:: python
+
+    # Run openmc in plotting mode
+    executor = openmc.Executor()
+    executor.plot_geometry(output=False)
+
+
+
+
+.. parsed-literal::
+
+    0
+
+
+
+.. code:: python
+
+    # Convert OpenMC's funky ppm to png
+    !convert materials-xy.ppm materials-xy.png
+    
+    # Display the materials plot inline
+    Image(filename='materials-xy.png')
+
+
+
+
+.. image:: post-processing-content_files/post-processing-content_24_0.png
+
+
+
+As we can see from the plot, we have a nice pin cell with fuel,
+cladding, and water! Before we run our simulation, we need to tell the
+code what we want to tally. The following code shows how to create a 2D
+mesh tally.
+
+.. code:: python
+
+    # Instantiate an empty TalliesFile
+    tallies_file = openmc.TalliesFile()
+
+.. code:: python
+
+    # Create mesh which will be used for tally
+    mesh = openmc.Mesh()
+    mesh.dimension = [100, 100]
+    mesh.lower_left = [-0.63, -0.63]
+    mesh.upper_right = [0.63, 0.63]
+    tallies_file.add_mesh(mesh)
+    
+    # Create mesh filter for tally
+    mesh_filter = openmc.Filter(type='mesh', bins=[1])
+    mesh_filter.mesh = mesh
+    
+    # Create mesh tally to score flux and fission rate
+    tally = openmc.Tally(name='flux')
+    tally.add_filter(mesh_filter)
+    tally.add_score('flux')
+    tally.add_score('fission')
+    tallies_file.add_tally(tally)
+
+.. code:: python
+
+    # Export to "tallies.xml"
+    tallies_file.export_to_xml()
+
+Now we a have a complete set of inputs, so we can go ahead and run our
+simulation.
+
+.. code:: python
+
+    # Run OpenMC!
+    executor.run_simulation()
+
+
+.. parsed-literal::
+
+    
+           .d88888b.                             888b     d888  .d8888b.
+          d88P" "Y88b                            8888b   d8888 d88P  Y88b
+          888     888                            88888b.d88888 888    888
+          888     888 88888b.   .d88b.  88888b.  888Y88888P888 888       
+          888     888 888 "88b d8P  Y8b 888 "88b 888 Y888P 888 888       
+          888     888 888  888 88888888 888  888 888  Y8P  888 888    888
+          Y88b. .d88P 888 d88P Y8b.     888  888 888   "   888 Y88b  d88P
+           "Y88888P"  88888P"   "Y8888  888  888 888       888  "Y8888P"
+    __________________888______________________________________________________
+                      888
+                      888
+    
+          Copyright:      2011-2015 Massachusetts Institute of Technology
+          License:        http://mit-crpg.github.io/openmc/license.html
+          Version:        0.7.0
+          Git SHA1:       c4b14a5ef87f004528d35cbf33fef3ed15a386ca
+          Date/Time:      2015-11-29 16:46:53
+          MPI Processes:  1
+    
+     ===========================================================================
+     ========================>     INITIALIZATION     <=========================
+     ===========================================================================
+    
+     Reading settings XML file...
+     Reading cross sections XML file...
+     Reading geometry XML file...
+     Reading materials XML file...
+     Reading tallies XML file...
+     Building neighboring cells lists for each surface...
+     Loading ACE cross section table: 92235.71c
+     Loading ACE cross section table: 92238.71c
+     Loading ACE cross section table: 8016.71c
+     Loading ACE cross section table: 1001.71c
+     Loading ACE cross section table: 5010.71c
+     Loading ACE cross section table: 40090.71c
+     Maximum neutron transport energy: 20.0000 MeV for 92235.71c
+     Initializing source particles...
+    
+     ===========================================================================
+     ====================>     K EIGENVALUE SIMULATION     <====================
+     ===========================================================================
+    
+      Bat./Gen.      k            Average k         
+      =========   ========   ====================   
+            1/1    1.04894                       
+            2/1    1.01711                       
+            3/1    1.05357                       
+            4/1    1.03052                       
+            5/1    1.06523                       
+            6/1    1.06806                       
+            7/1    1.05161                       
+            8/1    1.04199                       
+            9/1    1.05010                       
+           10/1    1.04617                       
+           11/1    1.04894                       
+           12/1    1.06806    1.05850 +/- 0.00956
+           13/1    1.05002    1.05567 +/- 0.00620
+           14/1    1.03471    1.05043 +/- 0.00683
+           15/1    1.01803    1.04395 +/- 0.00837
+           16/1    1.05588    1.04594 +/- 0.00712
+           17/1    1.07503    1.05010 +/- 0.00731
+           18/1    1.02786    1.04732 +/- 0.00691
+           19/1    1.00071    1.04214 +/- 0.00800
+           20/1    1.05587    1.04351 +/- 0.00729
+           21/1    1.03886    1.04309 +/- 0.00660
+           22/1    1.04335    1.04311 +/- 0.00603
+           23/1    1.04057    1.04292 +/- 0.00555
+           24/1    1.01976    1.04126 +/- 0.00540
+           25/1    1.05811    1.04238 +/- 0.00515
+           26/1    1.02351    1.04120 +/- 0.00496
+           27/1    1.05261    1.04188 +/- 0.00471
+           28/1    1.03355    1.04141 +/- 0.00446
+           29/1    1.02797    1.04071 +/- 0.00428
+           30/1    1.03758    1.04055 +/- 0.00406
+           31/1    1.04883    1.04094 +/- 0.00388
+           32/1    1.03557    1.04070 +/- 0.00371
+           33/1    1.02947    1.04021 +/- 0.00358
+           34/1    1.03651    1.04006 +/- 0.00343
+           35/1    1.03331    1.03979 +/- 0.00330
+           36/1    1.05947    1.04054 +/- 0.00326
+           37/1    1.05093    1.04093 +/- 0.00316
+           38/1    1.06787    1.04189 +/- 0.00319
+           39/1    1.01451    1.04095 +/- 0.00322
+           40/1    1.02351    1.04037 +/- 0.00317
+           41/1    1.04826    1.04062 +/- 0.00307
+           42/1    1.04228    1.04067 +/- 0.00298
+           43/1    1.03214    1.04041 +/- 0.00290
+           44/1    1.04950    1.04068 +/- 0.00282
+           45/1    1.06616    1.04141 +/- 0.00284
+           46/1    1.07039    1.04221 +/- 0.00287
+           47/1    1.00292    1.04115 +/- 0.00299
+           48/1    1.04477    1.04125 +/- 0.00291
+           49/1    1.03360    1.04105 +/- 0.00284
+           50/1    1.04783    1.04122 +/- 0.00277
+           51/1    1.03985    1.04119 +/- 0.00271
+           52/1    1.02507    1.04080 +/- 0.00267
+           53/1    1.03477    1.04066 +/- 0.00261
+           54/1    1.00412    1.03983 +/- 0.00268
+           55/1    1.02239    1.03945 +/- 0.00265
+           56/1    1.04308    1.03952 +/- 0.00259
+           57/1    1.05534    1.03986 +/- 0.00256
+           58/1    1.06667    1.04042 +/- 0.00257
+           59/1    1.06458    1.04091 +/- 0.00256
+           60/1    1.00304    1.04015 +/- 0.00262
+           61/1    1.05038    1.04036 +/- 0.00258
+           62/1    1.02904    1.04014 +/- 0.00254
+           63/1    1.00249    1.03943 +/- 0.00259
+           64/1    1.01779    1.03903 +/- 0.00257
+           65/1    1.05335    1.03929 +/- 0.00254
+           66/1    1.06231    1.03970 +/- 0.00253
+           67/1    1.02382    1.03942 +/- 0.00250
+           68/1    1.03796    1.03939 +/- 0.00245
+           69/1    1.03672    1.03935 +/- 0.00241
+           70/1    1.02926    1.03918 +/- 0.00238
+           71/1    1.05834    1.03950 +/- 0.00236
+           72/1    1.04332    1.03956 +/- 0.00232
+           73/1    1.05613    1.03982 +/- 0.00230
+           74/1    1.01963    1.03950 +/- 0.00228
+           75/1    1.02228    1.03924 +/- 0.00226
+           76/1    1.04842    1.03938 +/- 0.00223
+           77/1    1.02157    1.03911 +/- 0.00222
+           78/1    1.02810    1.03895 +/- 0.00219
+           79/1    1.05030    1.03912 +/- 0.00216
+           80/1    1.02391    1.03890 +/- 0.00214
+           81/1    1.02488    1.03870 +/- 0.00212
+           82/1    1.04957    1.03885 +/- 0.00210
+           83/1    1.03499    1.03880 +/- 0.00207
+           84/1    1.05922    1.03907 +/- 0.00206
+           85/1    1.05898    1.03934 +/- 0.00205
+           86/1    1.02242    1.03912 +/- 0.00204
+           87/1    1.03278    1.03904 +/- 0.00201
+           88/1    1.06134    1.03932 +/- 0.00201
+           89/1    1.04521    1.03940 +/- 0.00198
+           90/1    1.04277    1.03944 +/- 0.00196
+           91/1    1.04214    1.03947 +/- 0.00193
+           92/1    1.05610    1.03967 +/- 0.00192
+           93/1    1.04531    1.03974 +/- 0.00190
+           94/1    1.01534    1.03945 +/- 0.00190
+           95/1    1.03971    1.03945 +/- 0.00187
+           96/1    1.07183    1.03983 +/- 0.00189
+           97/1    1.07214    1.04020 +/- 0.00191
+           98/1    1.03710    1.04017 +/- 0.00188
+           99/1    1.02532    1.04000 +/- 0.00187
+          100/1    1.03965    1.04000 +/- 0.00185
+     Creating state point statepoint.100.h5...
+    
+     ===========================================================================
+     ======================>     SIMULATION FINISHED     <======================
+     ===========================================================================
+    
+    
+     =======================>     TIMING STATISTICS     <=======================
+    
+     Total time for initialization     =  3.7900E-01 seconds
+       Reading cross sections          =  8.7000E-02 seconds
+     Total time in simulation          =  2.2064E+02 seconds
+       Time in transport only          =  2.2060E+02 seconds
+       Time in inactive batches        =  8.7100E+00 seconds
+       Time in active batches          =  2.1193E+02 seconds
+       Time synchronizing fission bank =  1.4000E-02 seconds
+         Sampling source sites         =  8.0000E-03 seconds
+         SEND/RECV source sites        =  2.0000E-03 seconds
+       Time accumulating tallies       =  1.3000E-02 seconds
+     Total time for finalization       =  1.6600E-01 seconds
+     Total time elapsed                =  2.2120E+02 seconds
+     Calculation Rate (inactive)       =  5740.53 neutrons/second
+     Calculation Rate (active)         =  2123.37 neutrons/second
+    
+     ============================>     RESULTS     <============================
+    
+     k-effective (Collision)     =  1.03912 +/-  0.00160
+     k-effective (Track-length)  =  1.04000 +/-  0.00185
+     k-effective (Absorption)    =  1.04240 +/-  0.00156
+     Combined k-effective        =  1.04078 +/-  0.00127
+     Leakage Fraction            =  0.00000 +/-  0.00000
+    
+
+
+
+
+.. parsed-literal::
+
+    0
+
+
+
+Tally Data Processing
+---------------------
+
+Our simulation ran successfully and created a statepoint file with all
+the tally data in it. We begin our analysis here loading the statepoint
+file and 'reading' the results. By default, data from the statepoint
+file is only read into memory when it is requested. This helps keep the
+memory use to a minimum even when a statepoint file may be huge.
+
+.. code:: python
+
+    # Load the statepoint file
+    sp = StatePoint('statepoint.100.h5')
+
+Next we need to get the tally, which can be done with the
+``StatePoint.get_tally(...)`` method.
+
+.. code:: python
+
+    tally = sp.get_tally(scores=['flux'])
+    print(tally)
+
+
+.. parsed-literal::
+
+    Tally
+    	ID             =	10000
+    	Name           =	
+    	Filters        =	
+                    		mesh	[10000]
+    	Nuclides       =	total 
+    	Scores         =	[u'flux', u'fission']
+    	Estimator      =	tracklength
+    
+
+
+The statepoint file actually stores the sum and sum-of-squares for each
+tally bin from which the mean and variance can be calculated as
+described
+`here <http://mit-crpg.github.io/openmc/methods/tallies.html#variance>`__.
+The sum and sum-of-squares can be accessed using the ``sum`` and
+``sum_sq`` properties:
+
+.. code:: python
+
+    tally.sum
+
+
+
+
+.. parsed-literal::
+
+    array([[[ 0.4107676 ,  0.        ]],
+    
+           [[ 0.40849402,  0.        ]],
+    
+           [[ 0.41014343,  0.        ]],
+    
+           ..., 
+           [[ 0.41049467,  0.        ]],
+    
+           [[ 0.40982242,  0.        ]],
+    
+           [[ 0.40996987,  0.        ]]])
+
+
+
+However, the mean and standard deviation of the mean are usually what
+you are more interested in. The Tally class also has properties ``mean``
+and ``std_dev`` which automatically calculate these statistics
+on-the-fly.
+
+.. code:: python
+
+    print(tally.mean.shape)
+    (tally.mean, tally.std_dev)
+
+
+.. parsed-literal::
+
+    (10000, 1, 2)
+
+
+
+
+.. parsed-literal::
+
+    (array([[[ 0.00456408,  0.        ]],
+     
+            [[ 0.00453882,  0.        ]],
+     
+            [[ 0.00455715,  0.        ]],
+     
+            ..., 
+            [[ 0.00456105,  0.        ]],
+     
+            [[ 0.00455358,  0.        ]],
+     
+            [[ 0.00455522,  0.        ]]]),
+     array([[[  1.95085625e-05,   0.00000000e+00]],
+     
+            [[  1.78129859e-05,   0.00000000e+00]],
+     
+            [[  1.89709648e-05,   0.00000000e+00]],
+     
+            ..., 
+            [[  1.56286612e-05,   0.00000000e+00]],
+     
+            [[  1.65813279e-05,   0.00000000e+00]],
+     
+            [[  1.67530331e-05,   0.00000000e+00]]]))
+
+
+
+The tally data has three dimensions: one for filter combinations, one
+for nuclides, and one for scores. We see that there are 10000 filter
+combinations (corresponding to the 100 x 100 mesh bins), a single
+nuclide (since none was specified), and two scores. If we only want to
+look at a single score, we can use the ``get_slice(...)`` method as
+follows.
+
+.. code:: python
+
+    flux = tally.get_slice(scores=['flux'])
+    fission = tally.get_slice(scores=['fission'])
+    print(flux)
+
+
+.. parsed-literal::
+
+    Tally
+    	ID             =	10000
+    	Name           =	
+    	Filters        =	
+                    		mesh	[10000]
+    	Nuclides       =	total 
+    	Scores         =	[u'flux']
+    	Estimator      =	tracklength
+    
+
+
+To get the bins into a form that we can plot, we can simply change the
+shape of the array since it is a numpy array.
+
+.. code:: python
+
+    flux.std_dev.shape = (100, 100)
+    flux.mean.shape = (100, 100)
+    fission.std_dev.shape = (100, 100)
+    fission.mean.shape = (100, 100)
+
+.. code:: python
+
+    fig = plt.subplot(121)
+    fig.imshow(flux.mean)
+    fig2 = plt.subplot(122)
+    fig2.imshow(fission.mean)
+
+
+
+
+.. parsed-literal::
+
+    <matplotlib.image.AxesImage at 0x7fe38b3027d0>
+
+
+
+
+.. image:: post-processing-content_files/post-processing-content_44_1.png
+
+
+Now let's say we want to look at the distribution of relative errors of
+our tally bins for flux. First we create a new variable called
+``relative_error`` and set it to the ratio of the standard deviation and
+the mean, being careful not to divide by zero in case some bins were
+never scored to.
+
+.. code:: python
+
+    # Determine relative error
+    relative_error = np.zeros_like(flux.std_dev)
+    nonzero = flux.mean > 0
+    relative_error[nonzero] = flux.std_dev[nonzero] / flux.mean[nonzero]
+    
+    # distribution of relative errors
+    ret = plt.hist(relative_error[nonzero], bins=50)
+
+
+
+.. image:: post-processing-content_files/post-processing-content_46_0.png
+
+
+Source Sites
+------------
+
+Source sites can be accessed from the ``source`` property. As shown
+below, the source sites are represented as a numpy array with a
+structured datatype.
+
+.. code:: python
+
+    sp.source
+
+
+
+
+.. parsed-literal::
+
+    array([ (1.0, [0.2712169917165897, -0.04844236597355761, -0.1887902218343974], [0.3889598463000694, 0.8470657529949065, 0.36220139158953857], 2.2746035619924734, 0),
+           (1.0, [0.080729018085932, 0.19838688738571317, -0.38053428394017363], [-0.6604834049157511, -0.6893239101986768, 0.2976478097673534], 0.7833467555325838, 0),
+           (1.0, [0.019430574216787868, 0.06594180627832635, 0.23329810254580194], [-0.7472138923667574, 0.13227244377548197, -0.651287493870243], 1.1632342240714935, 0),
+           ...,
+           (1.0, [0.18544614514351207, -0.0113070561851496, 0.5468392238881264], [-0.8006491411918817, 0.43855795172388223, -0.4082007786475368], 1.4358240241589555, 0),
+           (1.0, [0.18544614514351207, -0.0113070561851496, 0.5468392238881264], [-0.5150076397044656, -0.34922134026850293, 0.7828228321575105], 1.5771133724329802, 0),
+           (1.0, [-0.2722999793764598, 0.22680062445008103, 0.2987060438567475], [0.9207818175032396, -0.2884020326181676, 0.26265017063984586], 2.932342523379745, 0)], 
+          dtype=[('wgt', '<f8'), ('xyz', '<f8', (3,)), ('uvw', '<f8', (3,)), ('E', '<f8'), ('delayed_group', '<i4')])
+
+
+
+If we want, say, only the energies from the source sites, we can simply
+index the source array with the name of the field:
+
+.. code:: python
+
+    sp.source['E']
+
+
+
+
+.. parsed-literal::
+
+    array([ 2.27460356,  0.78334676,  1.16323422, ...,  1.43582402,
+            1.57711337,  2.93234252])
+
+
+
+Now, we can look at things like the energy distribution of source sites.
+Note that we don't directly use the ``matplotlib.pyplot.hist`` method
+since our binning is logarithmic.
+
+.. code:: python
+
+    # Create log-spaced energy bins from 1 keV to 100 MeV
+    energy_bins = np.logspace(-3,1)
+    
+    # Calculate pdf for source energies
+    probability, bin_edges = np.histogram(sp.source['E'], energy_bins, density=True)
+    
+    # Make sure integrating the PDF gives us unity
+    print(sum(probability*np.diff(energy_bins)))
+    
+    # Plot source energy PDF
+    plt.semilogx(energy_bins[:-1], probability*np.diff(energy_bins), linestyle='steps')
+    plt.xlabel('Energy (MeV)')
+    plt.ylabel('Probability/MeV')
+
+
+.. parsed-literal::
+
+    1.0
+
+
+
+
+.. parsed-literal::
+
+    <matplotlib.text.Text at 0x7fe3652fd810>
+
+
+
+
+.. image:: post-processing-content_files/post-processing-content_53_2.png
+
+
+Let's also look at the spatial distribution of the sites. To make the
+plot a little more interesting, we can also include the direction of the
+particle emitted from the source and color each source by the logarithm
+of its energy.
+
+.. code:: python
+
+    plt.quiver(sp.source['xyz'][:,0], sp.source['xyz'][:,1],
+               sp.source['uvw'][:,0], sp.source['uvw'][:,1],
+               np.log(sp.source['E']), cmap='jet', scale=20.0)
+    plt.colorbar()
+    plt.xlim((-0.5,0.5))
+    plt.ylim((-0.5,0.5))
+
+
+
+
+.. parsed-literal::
+
+    (-0.5, 0.5)
+
+
+
+.. parsed-literal::
+
+    /usr/lib/pymodules/python2.7/matplotlib/collections.py:548: FutureWarning: elementwise comparison failed; returning scalar instead, but in the future will perform elementwise comparison
+      if self._edgecolors == 'face':
+
+
+
+.. image:: post-processing-content_files/post-processing-content_55_2.png
+

_sources/pythonapi/examples/post-processing.txt

@@ -0,0 +1,13 @@
+.. _notebook_post_processing:
+
+===============
+Post Processing
+===============
+
+.. only:: html
+
+    .. notebook:: post-processing.ipynb
+
+.. only:: latex
+
+    IPython notebooks must be viewed in the online HTML documentation.

_sources/pythonapi/index.txt

@@ -57,13 +57,26 @@ on a given module or class.
     summary
     tallies
 
+**Multi-Group Cross Section Generation**
+
+.. toctree::
+    :maxdepth: 1
+
+    mgxs
+    energy_groups
+    mgxs_library
+
 **Example Jupyter Notebooks:**
 
 .. toctree::
     :maxdepth: 1
 
+    examples/post-processing
     examples/pandas-dataframes
     examples/tally-arithmetic
+    examples/mgxs-part-i
+    examples/mgxs-part-ii
+    examples/mgxs-part-iii
 
 .. _Jupyter: https://jupyter.org/
 .. _NumPy: http://www.numpy.org/

_sources/pythonapi/mgxs.txt

@@ -0,0 +1,66 @@
+.. _pythonapi_mgxs:
+
+==========================
+Multi-Group Cross Sections
+==========================
+
+.. currentmodule:: openmc.mgxs.mgxs
+
+----------------------------
+Summary of Available Classes
+----------------------------
+
+.. autosummary::
+
+   MGXS
+   AbsorptionXS
+   CaptureXS
+   Chi
+   FissionXS
+   NuFissionXS
+   NuScatterXS
+   NuScatterMatrixXS
+   ScatterXS
+   ScatterMatrixXS
+   TotalXS
+   TransportXS
+
+-------------------
+Class Documentation
+-------------------
+
+.. autoclass:: MGXS
+    :members:
+
+.. autoclass:: AbsorptionXS
+    :members:
+
+.. autoclass:: CaptureXS
+    :members:
+
+.. autoclass:: Chi
+    :members:
+
+.. autoclass:: FissionXS
+    :members:
+
+.. autoclass:: NuFissionXS
+    :members:
+
+.. autoclass:: NuScatterXS
+    :members:
+
+.. autoclass:: NuScatterMatrixXS
+    :members:
+
+.. autoclass:: ScatterXS
+    :members:
+
+.. autoclass:: ScatterMatrixXS
+    :members:
+
+.. autoclass:: TotalXS
+    :members:
+
+.. autoclass:: TransportXS
+    :members:

_sources/pythonapi/mgxs_library.txt

@@ -0,0 +1,8 @@
+.. _pythonapi_mgxs_library:
+
+============
+MGXS Library
+============
+
+.. automodule:: openmc.mgxs.library
+    :members:

_sources/quickinstall.txt

@@ -35,8 +35,8 @@ Installing from Source on Linux or Mac OS X
 -------------------------------------------
 
 All OpenMC source code is hosted on GitHub_. If you have git_, the gfortran_
-compiler, and CMake_ installed, you can download and install OpenMC be entering
-the following commands in a terminal:
+compiler, CMake_, and HDF5_ installed, you can download and install OpenMC be
+entering the following commands in a terminal:
 
 .. code-block:: sh
 

_sources/releasenotes.txt

@@ -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>`_

_sources/usersguide/index.txt

@@ -5,7 +5,7 @@ User's Guide
 ============
 
 Welcome to the OpenMC User's Guide!  This tutorial will guide you through the
-essential aspects of using OpenMC to perform neutronic simulations.
+essential aspects of using OpenMC to perform simulations.
 
 .. toctree::
     :numbered:
@@ -14,5 +14,6 @@ essential aspects of using OpenMC to perform neutronic simulations.
     beginners
     install
     input
+    output/index
     processing
     troubleshoot

_sources/usersguide/input.txt

@@ -79,14 +79,13 @@ Message                   Description
 [VALID]                   XML file matches RelaxNG.
 ========================  ===================================
 
-As an example, if OpenMC is installed in the directory
-``/opt/openmc/0.6.2`` and the current working directory is where
-OpenMC XML input files are located, they can be validated using
-the following command:
+As an example, if OpenMC is installed in the directory ``/opt/openmc/`` and the
+current working directory is where OpenMC XML input files are located, they can
+be validated using the following command:
 
 .. code-block:: bash
 
-   /opt/openmc/0.6.2/bin/xml_validate
+   /opt/openmc/bin/openmc-validate-xml
 
 --------------------------------------
 Settings Specification -- settings.xml
@@ -721,9 +720,8 @@ Geometry Specification -- geometry.xml
 The geometry in OpenMC is described using `constructive solid geometry`_ (CSG),
 also sometimes referred to as combinatorial geometry. CSG allows a user to
 create complex objects using Boolean operators on a set of simpler surfaces. In
-the geometry model, each unique closed volume in defined by its bounding
-surfaces. In OpenMC, most `quadratic surfaces`_ can be modeled and used as
-bounding surfaces.
+the geometry model, each unique volume is defined by its bounding surfaces. In
+OpenMC, most `quadratic surfaces`_ can be modeled and used as bounding surfaces.
 
 Every geometry.xml must have an XML declaration at the beginning of the file and
 a root element named geometry. Within the root element the user can define any
@@ -746,7 +744,7 @@ number of cells, surfaces, and lattices. Let us look at the following example:
         <id>1</id>
         <universe>0</universe>
         <material>1</material>
-        <surfaces>-1</surfaces>
+        <region>-1</region>
       </cell>
     </geometry>
 
@@ -764,7 +762,7 @@ could be written as:
       <!-- This is a comment -->
 
       <surface id="1" type="sphere" coeffs="0.0 0.0 0.0 5.0" boundary="vacuum" />
-      <cell id="1" universe="0" material="1" surfaces="-1" />
+      <cell id="1" universe="0" material="1" region="-1" />
 
     </geometry>
 
@@ -788,7 +786,8 @@ Each ``<surface>`` element can have the following attributes or sub-elements:
 
   :type:
     The type of the surfaces. This can be "x-plane", "y-plane", "z-plane",
-    "plane", "x-cylinder", "y-cylinder", "z-cylinder", or "sphere".
+    "plane", "x-cylinder", "y-cylinder", "z-cylinder", "sphere", "x-cone",
+    "y-cone", "z-cone", or "quadric".
 
     *Default*: None
 
@@ -856,6 +855,12 @@ The following quadratic surfaces can be modeled:
     R^2 (z - z_0)^2`. The coefficients specified are ":math:`x_0 \: y_0 \: z_0
     \: R^2`".
 
+  :quadric:
+     A general quadric surface of the form :math:`Ax^2 + By^2 + Cz^2 + Dxy +
+     Eyz + Fxz + Gx + Hy + Jz + K = 0` The coefficients specified are ":math:`A
+     \: B \: C \: D \: E \: F \: G \: H \: J \: K`".
+
+
 ``<cell>`` Element
 ------------------
 
@@ -892,15 +897,29 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
 
     *Default*: None
 
-  :surfaces:
-    A list of the ``ids`` for surfaces that bound this cell, e.g. if the cell
-    is on the negative side of surface 3 and the positive side of surface 5, the
-    bounding surfaces would be given as "-3 5".
+  :region:
+    A Boolean expression of half-spaces that defines the spatial region which
+    the cell occupies. Each half-space is identified by the unique ID of the
+    surface prefixed by `-` or `+` to indicate that it is the negative or
+    positive half-space, respectively. The `+` sign for a positive half-space
+    can be omitted. Valid Boolean operators are parentheses, union `|`,
+    complement `~`, and intersection. Intersection is implicit and indicated by
+    the presence of whitespace. The order of operator precedence is parentheses,
+    complement, intersection, and then union.
 
-    .. note:: The surface attribute/element can be omitted to make a cell fill
-              its entire universe.
+    As an example, the following code gives a cell that is the union of the
+    negative half-space of surface 3 and the complement of the intersection of
+    the positive half-space of surface 5 and the negative half-space of surface
+    2:
 
-    *Default*: No surfaces
+    .. code-block:: xml
+
+        <cell id="1" material="1" region="-3 | ~(5 -2)" />
+
+    .. note:: The ``region`` attribute/element can be omitted to make a cell
+              fill its entire universe.
+
+    *Default*: A region filling all space.
 
   :rotation:
     If the cell is filled with a universe, this element specifies the angles in
@@ -1095,8 +1114,10 @@ Each ``material`` element can have the following attributes or sub-elements:
     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", or "sum". 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
+    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.
 
     *Default*: None
@@ -1117,6 +1138,15 @@ Each ``material`` element can have the following attributes or sub-elements:
     .. note:: If one nuclide is specified in atom percent, all others must also
               be given in atom percent. The same applies for weight percentages.
 
+    An optional attribute/sub-element for each nuclide is ``scattering``. This
+    attribute may be set to "data" to use the scattering laws specified by the
+    cross section library (default). Alternatively, when set to "iso-in-lab",
+    the scattering laws are used to sample the outgoing energy but an
+    isotropic-in-lab  distribution is used to sample the outgoing angle at each
+    scattering interaction. The ``scattering`` attribute may be most useful
+    when using OpenMC to compute multi-group cross-sections for deterministic
+    transport codes and to quantify the effects of anisotropic scattering.
+
     *Default*: None
 
   :element:
@@ -1143,6 +1173,16 @@ Each ``material`` element can have the following attributes or sub-elements:
 
     *Default*: None
 
+    An optional attribute/sub-element for each element is ``scattering``. This
+    attribute may be set to "data" to use the scattering laws specified by the
+    cross section library (default). Alternatively, when set to "iso-in-lab",
+    the scattering laws are used to sample the outgoing energy but an
+    isotropic-in-lab  distribution is used to sample the outgoing angle at each
+    scattering interaction. The ``scattering`` attribute may be most useful
+    when using OpenMC to compute multi-group cross-sections for deterministic
+    transport codes and to quantify the effects of anisotropic scattering.
+
+    *Default*: None
 
   :sab:
     Associates an S(a,b) table with the material. This element has
@@ -1214,8 +1254,8 @@ The ``<tally>`` element accepts the following sub-elements:
 
       :type:
         The type of the filter. Accepted options are "cell", "cellborn",
-        "material", "universe", "energy", "energyout", "mesh", and
-        "distribcell".
+        "material", "universe", "energy", "energyout", "mesh", "distribcell",
+        and "delayedgroup".
 
       :bins:
         For each filter type, the corresponding ``bins`` entry is given as
@@ -1240,17 +1280,87 @@ The ``<tally>`` element accepts the following sub-elements:
         :energy:
           A monotonically increasing list of bounding **pre-collision** energies
           for a number of groups. For example, if this filter is specified as
-          ``<filter type="energy" bins="0.0 1.0 20.0" />``, then two energy bins
-          will be created, one with energies between 0 and 1 MeV and the other
-          with energies between 1 and 20 MeV.
+
+          .. code-block:: xml
+
+              <filter type="energy" bins="0.0 1.0 20.0" />
+
+          then two energy bins will be created, one with energies between 0 and
+          1 MeV and the other with energies between 1 and 20 MeV.
 
         :energyout:
           A monotonically increasing list of bounding **post-collision**
           energies for a number of groups. For example, if this filter is
-          specified as ``<filter type="energyout" bins="0.0 1.0 20.0" />``, then
-          two post-collision energy bins will be created, one with energies
+          specified as
+
+          .. code-block:: xml
+
+              <filter type="energyout" bins="0.0 1.0 20.0" />
+
+          then two post-collision energy bins will be created, one with energies
           between 0 and 1 MeV and the other with energies between 1 and 20 MeV.
 
+        :mu:
+          A monotonically increasing list of bounding **post-collision** cosines
+          of the change in a particle's angle (i.e., :math:`\mu = \hat{\Omega}
+          \cdot \hat{\Omega}'`), which represents a portion of the possible
+          values of :math:`[-1,1]`.  For example, spanning all of :math:`[-1,1]`
+          with five equi-width bins can be specified as:
+
+          .. code-block:: xml
+
+              <filter type="mu" bins="-1.0 -0.6 -0.2 0.2 0.6 1.0" />
+
+          Alternatively, if only one value is provided as a bin, OpenMC will
+          interpret this to mean the complete range of :math:`[-1,1]` should
+          be automatically subdivided in to the provided value for the bin.
+          That is, the above example of five equi-width bins spanning
+          :math:`[-1,1]` can be instead written as:
+
+          .. code-block:: xml
+
+              <filter type="mu" bins="5" />
+
+        :polar:
+          A monotonically increasing list of bounding particle polar angles
+          which represents a portion of the possible values of :math:`[0,\pi]`.
+          For example, spanning all of :math:`[0,\pi]` with five equi-width
+          bins can be specified as:
+
+          .. code-block:: xml
+
+              <filter type="polar" bins="0.0 0.6283 1.2566 1.8850 2.5132 3.1416"/>
+
+          Alternatively, if only one value is provided as a bin, OpenMC will
+          interpret this to mean the complete range of :math:`[0,\pi]` should
+          be automatically subdivided in to the provided value for the bin.
+          That is, the above example of five equi-width bins spanning
+          :math:`[0,\pi]` can be instead written as:
+
+          .. code-block:: xml
+
+              <filter type="polar" bins="5" />
+
+        :azimuthal:
+          A monotonically increasing list of bounding particle azimuthal angles
+          which represents a portion of the possible values of :math:`[-\pi,\pi)`.
+          For example, spanning all of :math:`[-\pi,\pi)` with two equi-width
+          bins can be specified as:
+
+          .. code-block:: xml
+
+              <filter type="azimuthal" bins="0.0 3.1416 6.2832" />
+
+          Alternatively, if only one value is provided as a bin, OpenMC will
+          interpret this to mean the complete range of :math:`[-\pi,\pi)` should
+          be automatically subdivided in to the provided value for the bin.
+          That is, the above example of five equi-width bins spanning
+          :math:`[-\pi,\pi)` can be instead written as:
+
+          .. code-block:: xml
+
+              <filter type="azimuthal" bins="2" />
+
         :mesh:
           The ``id`` of a structured mesh to be tallied over.
 
@@ -1263,6 +1373,15 @@ The ``<tally>`` element accepts the following sub-elements:
               not accept more than one cell ID. It is not recommended to combine
               this filter with a cell or mesh filter.
 
+        :delayedgroup:
+          A list of delayed neutron precursor groups for which the tally should
+          be accumulated. For instance, to tally to all 6 delayed groups in the
+          ENDF/B-VII.1 library the filter is specified as:
+
+          .. code-block:: xml
+
+              <filter type="delayedgroup" bins="1 2 3 4 5 6" />
+
   :nuclides:
     If specified, the scores listed will be for particular nuclides, not the
     summation of reactions from all nuclides. The format for nuclides should be
@@ -1278,26 +1397,32 @@ The ``<tally>`` element accepts the following sub-elements:
     *Default*: total
 
   :estimator:
-    The estimator element is used to force the use of either ``analog`` or
-    ``tracklength`` tally estimation.  ''analog'' is generally less efficient
-    though it can be used with every score type.  ''tracklength'' is generally
-    the most efficient, though its usage is restricted to tallies that do not
-    score particle information which requires a collision to have occured, such
-    as a scattering tally which utilizes outgoing energy filters.
+    The estimator element is used to force the use of either ``analog``,
+    ``collision``, or ``tracklength`` tally estimation.  ``analog`` is generally
+    the least efficient though it can be used with every score type.
+    ``tracklength`` is generally the most efficient, but neither ``tracklength``
+    nor ``collision`` can be used to score a tally that requires post-collision
+    information.  For example, a scattering tally with outgoing energy filters
+    cannot be used with ``tracklength`` or ``collision`` because the code will
+    not know the outgoing energy distribution.
 
-    *Default*: ``tracklength`` but will revert to analog if necessary.
+    *Default*: ``tracklength`` but will revert to ``analog`` if necessary.
 
   :scores:
     A space-separated list of the desired responses to be accumulated. Accepted
     options are "flux", "total", "scatter", "absorption", "fission",
-    "nu-fission", "kappa-fission", "nu-scatter", "scatter-N", "scatter-PN",
-    "scatter-YN", "nu-scatter-N", "nu-scatter-PN", "nu-scatter-YN", "flux-YN",
-    "total-YN", "current", and "events". These corresponding to the following
-    physical quantities:
+    "nu-fission", "delayed-nu-fission", "kappa-fission", "nu-scatter",
+    "scatter-N", "scatter-PN", "scatter-YN", "nu-scatter-N", "nu-scatter-PN",
+    "nu-scatter-YN", "flux-YN", "total-YN", "current", "inverse-velocity" and
+    "events". These correspond to the following physical quantities:
 
     :flux:
       Total flux in particle-cm per source particle.
 
+      .. note::
+         The ``analog`` estimator is actually identical to the ``collision``
+         estimator for the flux score.
+
     :total:
       Total reaction rate in reactions per source particle.
 
@@ -1316,6 +1441,10 @@ The ``<tally>`` element accepts the following sub-elements:
       Total production of neutrons due to fission. Units are neutrons produced
       per source neutron.
 
+    :delayed-nu-fission:
+      Total production of delayed neutrons due to fission. Units are neutrons produced
+      per source neutron.
+
     :kappa-fission:
       The recoverable energy production rate due to fission. The recoverable
       energy is defined as the fission product kinetic energy, prompt and
@@ -1378,6 +1507,14 @@ The ``<tally>`` element accepts the following sub-elements:
           specified. Furthermore, it may not be used in conjunction with any
           other score.
 
+    :inverse-velocity:
+      The flux-weighted inverse velocity where the velocity is in units of
+      centimeters per second.
+
+      .. note::
+         The ``analog`` estimator is actually identical to the ``collision``
+         estimator for the inverse-velocity score.
+
     :events:
       Number of scoring events. Units are events per source particle.
 
@@ -1423,8 +1560,7 @@ a separate element with the tag name ``<mesh>``. This element has the following
 attributes/sub-elements:
 
   :type:
-    The type of structured mesh. Valid options include "rectangular" and
-    "hexagonal".
+    The type of structured mesh. The only valid option is "regular".
 
   :dimension:
     The number of mesh cells in each direction.
@@ -1526,16 +1662,16 @@ sub-elements:
     *Default*: None - Required entry
 
   :type:
-    Keyword for type of plot to be produced. Currently only "slice" and
-    "voxel" plots are implemented. The "slice" plot type creates 2D pixel
-    maps saved in the PPM file format. PPM files can be displayed in most
-    viewers (e.g. the default Gnome viewer, IrfanView, etc.).  The "voxel"
-    plot type produces a binary datafile containing voxel grid positioning and
-    the cell or material (specified by the ``color`` tag) at the center of each
-    voxel. These datafiles can be processed into 3D SILO files using the
-    ``voxel.py`` utility provided with the OpenMC source, and subsequently
-    viewed with a 3D viewer such as VISIT or Paraview. See the
-    :ref:`devguide_voxel` for information about the datafile structure.
+    Keyword for type of plot to be produced. Currently only "slice" and "voxel"
+    plots are implemented. The "slice" plot type creates 2D pixel maps saved in
+    the PPM file format. PPM files can be displayed in most viewers (e.g. the
+    default Gnome viewer, IrfanView, etc.).  The "voxel" plot type produces a
+    binary datafile containing voxel grid positioning and the cell or material
+    (specified by the ``color`` tag) at the center of each voxel. These
+    datafiles can be processed into 3D SILO files using the
+    ``openmc-voxel-to-silovtk`` utility provided with the OpenMC source, and
+    subsequently viewed with a 3D viewer such as VISIT or Paraview. See the
+    :ref:`usersguide_voxel` for information about the datafile structure.
 
     .. note:: Since the PPM format is saved without any kind of compression,
               the resulting file sizes can be quite large.  Saving the image in

_sources/usersguide/install.txt

@@ -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
 --------------------
@@ -59,6 +62,37 @@ Prerequisites
 
           sudo apt-get install cmake
 
+    * HDF5_ Library for portable binary output format
+
+      OpenMC uses HDF5 for binary output files. As such, you will need to have
+      HDF5 installed on your computer. The installed version will need to have
+      been compiled with the same compiler you intend to compile OpenMC with. If
+      you are using HDF5 in conjunction with MPI, we recommend that your HDF5
+      installation be built with parallel I/O features. An example of
+      configuring HDF5_ is listed below::
+
+           FC=/opt/mpich/3.1/bin/mpif90 CC=/opt/mpich/3.1/bin/mpicc \
+           ./configure --prefix=/opt/hdf5/1.8.12 --enable-fortran \
+                       --enable-fortran2003 --enable-parallel
+
+      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:
+
+      .. code-block:: sh
+
+          sudo apt-get install libhdf5-8 libhdf5-dev hdf5-helpers
+
+      Note that the exact package names may vary depending on your particular
+      distribution and version.
+
 .. admonition:: Optional
 
     * An MPI implementation for distributed-memory parallel runs
@@ -72,20 +106,6 @@ Prerequisites
           sudo apt-get install mpich libmpich-dev
           sudo apt-get install openmpi-bin libopenmpi1.6 libopenmpi-dev
 
-    * HDF5_ Library for portable binary output format
-
-      To compile with support for HDF5_ output (highly recommended), you will
-      need to have HDF5 installed on your computer. The installed version will
-      need to have been compiled with the same compiler you intend to compile
-      OpenMC with. HDF5_ must be built with parallel I/O features if you intend
-      to use HDF5_ with MPI. An example of configuring HDF5_ is listed below::
-
-           FC=/opt/mpich/3.1/bin/mpif90 CC=/opt/mpich/3.1/bin/mpicc \
-           ./configure --prefix=/opt/hdf5/1.8.12 --enable-fortran \
-                       --enable-fortran2003 --enable-parallel
-
-      You may omit ``--enable-parallel`` if you want to compile HDF5_ in serial.
-
     * git_ version control software for obtaining source code
 
 .. _gfortran: http://gcc.gnu.org/wiki/GFortran
@@ -194,27 +214,26 @@ command, i.e.
 
     FC=mpif90 cmake /path/to/openmc
 
-Compiling with HDF5
-+++++++++++++++++++
-
-To compile with MPI, set the :envvar:`FC` environment variable to the path to
-the HDF5 Fortran wrapper. For example, in a bash shell:
+Selecting HDF5 Installation
++++++++++++++++++++++++++++
 
+CMakeLists.txt searches for the ``h5fc`` or ``h5pfc`` HDF5 Fortran wrapper on
+your PATH environment variable and subsequently uses it to determine library
+locations and compile flags. If you have multiple installations of HDF5 or one
+that does not appear on your PATH, you can set the HDF5_ROOT environment
+variable to the root directory of the HDF5 installation, e.g.
 .. code-block:: sh
 
-    export FC=h5fc
+    export HDF5_ROOT=/opt/hdf5/1.8.15
     cmake /path/to/openmc
 
-As noted above, an environment variable can typically be set for a single
-command, i.e.
+This will cause CMake to search first in /opt/hdf5/1.8.15/bin for ``h5fc`` /
+``h5pfc`` before it searches elsewhere. As noted above, an environment variable
+can typically be set for a single command, i.e.
 
 .. code-block:: sh
 
-    FC=h5fc cmake /path/to/openmc
-
-To compile with support for both MPI and HDF5, use the parallel HDF5 wrapper
-``h5pfc`` instead. Note that this requires that your HDF5 installation be
-compiled with ``--enable-parallel``.
+    HDF5_ROOT=/opt/hdf5/1.8.15 cmake /path/to/openmc
 
 Compiling on Linux and Mac OS X
 -------------------------------
@@ -308,6 +327,25 @@ This will build an executable named ``openmc``.
 .. _MinGW: http://www.mingw.org
 .. _SourceForge: http://sourceforge.net/projects/mingw
 
+Compiling for the Intel Xeon Phi
+--------------------------------
+
+In order to build OpenMC for the Intel Xeon Phi using the Intel Fortran
+compiler, it is necessary to specify that all objects be compiled with the
+``-mmic`` flag as follows:
+
+.. code-block:: sh
+
+    mkdir build && cd build
+    FC=ifort FFLAGS=-mmic cmake -Dopenmp=on ..
+    make
+
+Note that unless an HDF5 build for the Intel Xeon Phi is already on your target
+machine, you will need to cross-compile HDF5 for the Xeon Phi. An `example
+script`_ to build zlib and HDF5 provides several necessary workarounds.
+
+.. _example script: https://github.com/paulromano/install-scripts/blob/master/install-hdf5-mic
+
 Testing Build
 -------------
 

_sources/usersguide/output/index.txt

@@ -0,0 +1,16 @@
+.. _usersguide_output:
+
+===================
+Output File Formats
+===================
+
+.. toctree::
+   :numbered:
+   :maxdepth: 3
+
+   statepoint
+   source
+   summary
+   particle_restart
+   track
+   voxel

_sources/usersguide/output/particle_restart.txt

@@ -0,0 +1,57 @@
+.. _usersguide_particle_restart:
+
+============================
+Particle Restart File Format
+============================
+
+The current revision of the particle restart file format is 1.
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/revision** (*int*)
+
+    Revision of the particle restart file format. Any time a change is made in
+    the format, this integer is incremented.
+
+**/current_batch** (*int*)
+
+    The number of batches already simulated.
+
+**/gen_per_batch** (*int*)
+
+    Number of generations per batch.
+
+**/current_gen** (*int*)
+
+    The number of generations already simulated.
+
+**/n_particles** (*int8_t*)
+
+    Number of particles used per generation.
+
+**/run_mode** (*int*)
+
+    Run mode used. A value of 1 indicates a fixed-source run and a value of 2
+    indicates an eigenvalue run.
+
+**/id** (*int8_t*)
+
+    Unique identifier of the particle.
+
+**/weight** (*double*)
+
+    Weight of the particle.
+
+**/energy** (*double*)
+
+    Energy of the particle in MeV.
+
+**/xyz** (*double[3]*)
+
+    Position of the particle.
+
+**/uvw** (*double[3]*)
+
+    Direction of the particle.

_sources/usersguide/output/source.txt

@@ -0,0 +1,19 @@
+.. _usersguide_source:
+
+==================
+Source File Format
+==================
+
+Normally, source data is stored in a state point file. However, it is possible
+to request that the source be written separately, in which case the format used
+is that documented here.
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/source_bank** (Compound type)
+
+    Source bank information for each particle. The compound type has fields
+    ``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight, position,
+    direction, and energy of the source particle, respectively.

_sources/usersguide/output/statepoint.txt

@@ -0,0 +1,259 @@
+.. _usersguide_statepoint:
+
+=======================
+State Point File Format
+=======================
+
+The current revision of the statepoint file format is 14.
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/revision** (*int*)
+
+    Revision of the state point file format. Any time a change is made in the
+    format, this integer is incremented.
+
+**/version_major** (*int*)
+
+    Major version number for OpenMC
+
+**/version_minor** (*int*)
+
+    Minor version number for OpenMC
+
+**/version_release** (*int*)
+
+    Release version number for OpenMC
+
+**/date_and_time** (*char[]*)
+
+    Date and time the state point was written.
+
+**/path** (*char[]*)
+
+    Absolute path to directory containing input files.
+
+**/seed** (*int8_t*)
+
+    Pseudo-random number generator seed.
+
+**/run_mode** (*char[]*)
+
+    Run mode used. A value of 1 indicates a fixed-source run and a value of 2
+    indicates an eigenvalue run.
+
+**/n_particles** (*int8_t*)
+
+    Number of particles used per generation.
+
+**/n_batches** (*int*)
+
+    Number of batches to simulate.
+
+**/current_batch** (*int*)
+
+    The number of batches already simulated.
+
+if run_mode == 'k-eigenvalue':
+
+    **/n_inactive** (*int*)
+
+        Number of inactive batches.
+
+    **/gen_per_batch** (*int*)
+
+        Number of generations per batch.
+
+    **/k_generation** (*double[]*)
+
+        k-effective for each generation simulated.
+
+    **/entropy** (*double[]*)
+
+        Shannon entropy for each generation simulated
+
+    **/k_col_abs** (*double*)
+
+        Sum of product of collision/absorption estimates of k-effective
+
+    **/k_col_tra** (*double*)
+
+        Sum of product of collision/track-length estimates of k-effective
+
+    **/k_abs_tra** (*double*)
+
+        Sum of product of absorption/track-length estimates of k-effective
+
+    **/k_combined** (*double[2]*)
+
+        Mean and standard deviation of a combined estimate of k-effective
+
+    **/cmfd_on** (*int*)
+
+        Flag indicating whether CMFD is on (1) or off (0).
+
+    if (cmfd_on)
+
+        **/cmfd/indices** (*int[4]*)
+
+            Indices for cmfd mesh (i,j,k,g)
+
+        **/cmfd/k_cmfd** (*double[]*)
+
+            CMFD eigenvalues
+
+        **/cmfd/cmfd_src** (*double[][][][]*)
+
+            CMFD fission source
+
+        **/cmfd/cmfd_entropy** (*double[]*)
+
+            CMFD estimate of Shannon entropy
+
+        **/cmfd/cmfd_balance** (*double[]*)
+
+            RMS of the residual neutron balance equation on CMFD mesh
+
+        **/cmfd/cmfd_dominance** (*double[]*)
+
+            CMFD estimate of dominance ratio
+
+        **/cmfd/cmfd_srccmp** (*double[]*)
+
+            RMS comparison of difference between OpenMC and CMFD fission source
+
+**/tallies/n_meshes** (*int*)
+
+    Number of meshes in tallies.xml file
+
+**/tally/meshes/ids** (*int[]*)
+
+    Internal unique ID of each mesh.
+
+**/tally/meshes/keys** (*int[]*)
+
+    User-identified unique ID of each mesh.
+
+**/tallies/meshes/mesh <uid>/type** (*char[]*)
+
+    Type of mesh.
+
+**/tallies/meshes/mesh <uid>/dimension** (*int*)
+
+    Number of mesh cells in each dimension.
+
+**/tallies/meshes/mesh <uid>/lower_left** (*double[]*)
+
+    Coordinates of lower-left corner of mesh.
+
+**/tallies/meshes/mesh <uid>/upper_right** (*double[]*)
+
+    Coordinates of upper-right corner of mesh.
+
+**/tallies/meshes/mesh <uid>/width** (*double[]*)
+
+    Width of each mesh cell in each dimension.
+
+**/tallies/n_tallies** (*int*)
+
+    Number of user-defined tallies.
+
+**/tallies/ids** (*int[]*)
+
+    Internal unique ID of each tally.
+
+**/tallies/keys** (*int[]*)
+
+    User-identified unique ID of each tally.
+
+**/tallies/tally <uid>/estimator** (*char[]*)
+
+    Type of tally estimator, either 'analog', 'tracklength', or 'collision'.
+
+**/tallies/tally <uid>/n_realizations** (*int*)
+
+    Number of realizations.
+
+**/tallies/tally <uid>/n_filters** (*int*)
+
+    Number of filters used.
+
+**/tallies/tally <uid>/filter <j>/type** (*char[]*)
+
+    Type of the j-th filter. Can be 'universe', 'material', 'cell', 'cellborn',
+    'surface', 'mesh', 'energy', 'energyout', or 'distribcell'.
+
+**/tallies/tally <uid>/filter <j>/offset** (*int*)
+
+    Filter offset (used for distribcell filter).
+
+**/tallies/tally <uid>/filter <j>/n_bins** (*int*)
+
+    Number of bins for the j-th filter.
+
+**/tallies/tally <uid>/filter <j>/bins** (*int[]* or *double[]*)
+
+    Value for each filter bin of this type.
+
+**/tallies/tally <uid>/nuclides** (*char[][]*)
+
+    Array of nuclides to tally. Note that if no nuclide is specified in the user
+    input, a single 'total' nuclide appears here.
+
+**/tallies/tally <uid>/n_score_bins** (*int*)
+
+    Number of scoring bins for a single nuclide. In general, this can be greater
+    than the number of user-specified scores since each score might have
+    multiple scoring bins, e.g., scatter-PN.
+
+**/tallies/tally <uid>/score_bins** (*char[][]*)
+
+    Values of specified scores.
+
+**/tallies/tally <uid>/n_user_scores** (*int*)
+
+    Number of scores without accounting for those added by expansions,
+    e.g. scatter-PN.
+
+**/tallies/tally <uid>/moment_orders** (*char[][]*)
+
+    Tallying moment orders for Legendre and spherical harmonic tally expansions
+    (*e.g.*, 'P2', 'Y1,2', etc.).
+
+**/tallies/tally <uid>/results** (Compound type)
+
+    Accumulated sum and sum-of-squares for each bin of the i-th tally. This is a
+    two-dimensional array, the first dimension of which represents combinations
+    of filter bins and the second dimensions of which represents scoring
+    bins. Each element of the array has fields 'sum' and 'sum_sq'.
+
+**/source_present** (*int*)
+
+    Flag indicated if source bank is present in the file
+
+**/n_realizations** (*int*)
+
+    Number of realizations for global tallies.
+
+**/n_global_tallies** (*int*)
+
+    Number of global tally scores.
+
+**/global_tallies** (Compound type)
+
+    Accumulated sum and sum-of-squares for each global tally. The compound type
+    has fields named ``sum`` and ``sum_sq``.
+
+**tallies_present** (*int*)
+
+    Flag indicated if tallies are present in the file.
+
+if (run_mode == 'k-eigenvalue' and source_present > 0)
+
+    **/source_bank** (Compound type)
+
+        Source bank information for each particle. The compound type has fields
+        ``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight,
+        position, direction, and energy of the source particle, respectively.

_sources/usersguide/output/summary.txt

@@ -0,0 +1,311 @@
+.. _usersguide_summary:
+
+===================
+Summary File Format
+===================
+
+The current revision of the summary file format is 1.
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/revision** (*int*)
+
+    Revision of the summary file format. Any time a change is made in the
+    format, this integer is incremented.
+
+**/version_major** (*int*)
+
+    Major version number for OpenMC
+
+**/version_minor** (*int*)
+
+    Minor version number for OpenMC
+
+**/version_release** (*int*)
+
+    Release version number for OpenMC
+
+**/date_and_time** (*char[]*)
+
+    Date and time the summary was written.
+
+**/n_procs** (*int*)
+
+    Number of MPI processes used.
+
+**/n_particles** (*int8_t*)
+
+    Number of particles used per generation.
+
+**/n_batches** (*int*)
+
+    Number of batches to simulate.
+
+**/n_inactive** (*int*)
+
+    Number of inactive batches. Only present if /run_mode is set to
+    'k-eigenvalue'.
+
+**/n_active** (*int*)
+
+    Number of active batches. Only present if /run_mode is set to
+    'k-eigenvalue'.
+
+**/gen_per_batch** (*int*)
+
+    Number of generations per batch. Only present if /run_mode is set to
+    'k-eigenvalue'.
+
+**/geometry/n_cells** (*int*)
+
+    Number of cells in the problem.
+
+**/geometry/n_surfaces** (*int*)
+
+    Number of surfaces in the problem.
+
+**/geometry/n_universes** (*int*)
+
+    Number of unique universes in the problem.
+
+**/geometry/n_lattices** (*int*)
+
+    Number of lattices in the problem.
+
+**/geometry/cells/cell <uid>/index** (*int*)
+
+    Index in cells array used internally in OpenMC.
+
+**/geometry/cells/cell <uid>/name** (*char[]*)
+
+    Name of the cell.
+
+**/geometry/cells/cell <uid>/universe** (*int*)
+
+    Universe assigned to the cell. If none is specified, the default
+    universe (0) is assigned.
+
+**/geometry/cells/cell <uid>/fill_type** (*char[]*)
+
+    Type of fill for the cell. Can be 'normal', 'universe', or 'lattice'.
+
+**/geometry/cells/cell <uid>/material** (*int*)
+
+    Unique ID of the material assigned to the cell. This dataset is present only
+    if fill_type is set to 'normal'.
+
+**/geometry/cells/cell <uid>/offset** (*int[]*)
+
+    Offsets used for distribcell tally filter. This dataset is present only if
+    fill_type is set to 'universe'.
+
+**/geometry/cells/cell <uid>/translation** (*double[3]*)
+
+    Translation applied to the fill universe. This dataset is present only if
+    fill_type is set to 'universe'.
+
+**/geometry/cells/cell <uid>/rotation** (*double[3]*)
+
+    Angles in degrees about the x-, y-, and z-axes for which the fill universe
+    should be rotated. This dataset is present only if fill_type is set to
+    'universe'.
+
+**/geometry/cells/cell <uid>/lattice** (*int*)
+
+    Unique ID of the lattice which fills the cell. Only present if fill_type is
+    set to 'lattice'.
+
+**/geometry/cells/cell <uid>/region** (*char[]*)
+
+    Region specification for the cell.
+
+**/geometry/surfaces/surface <uid>/index** (*int*)
+
+    Index in surfaces array used internally in OpenMC.
+
+**/geometry/surfaces/surface <uid>/name** (*char[]*)
+
+    Name of the surface.
+
+**/geometry/surfaces/surface <uid>/type** (*char[]*)
+
+    Type of the surface. Can be 'x-plane', 'y-plane', 'z-plane', 'plane',
+    'x-cylinder', 'y-cylinder', 'sphere', 'x-cone', 'y-cone', 'z-cone', or
+    'quadric'.
+
+**/geometry/surfaces/surface <uid>/coefficients** (*double[]*)
+
+    Array of coefficients that define the surface. See :ref:`surface_element`
+    for what coefficients are defined for each surface type.
+
+**/geometry/surfaces/surface <uid>/boundary_condition** (*char[]*)
+
+    Boundary condition applied to the surface. Can be 'transmission', 'vacuum',
+    'reflective', or 'periodic'.
+
+**/geometry/universes/universe <uid>/index** (*int*)
+
+    Index in the universes array used internally in OpenMC.
+
+**/geometry/universes/universe <uid>/cells** (*int[]*)
+
+    Array of unique IDs of cells that appear in the universe.
+
+**/geometry/lattices/lattice <uid>/index** (*int*)
+
+    Index in the lattices array used internally in OpenMC.
+
+**/geometry/lattices/lattice <uid>/name** (*char[]*)
+
+    Name of the lattice.
+
+**/geometry/lattices/lattice <uid>/type** (*char[]*)
+
+    Type of the lattice, either 'rectangular' or 'hexagonal'.
+
+**/geometry/lattices/lattice <uid>/pitch** (*double[]*)
+
+    Pitch of the lattice.
+
+**/geometry/lattices/lattice <uid>/outer** (*int*)
+
+    Outer universe assigned to lattice cells outside the defined range.
+
+**/geometry/lattices/lattice <uid>/offsets** (*int[]*)
+
+    Offsets used for distribcell tally filter.
+
+**/geometry/lattices/lattice <uid>/universes** (*int[]*)
+
+    Three-dimensional array of universes assigned to each cell of the lattice.
+
+**/geometry/lattices/lattice <uid>/dimension** (*int[]*)
+
+    The number of lattice cells in each direction. This dataset is present only
+    when the 'type' dataset is set to 'rectangular'.
+
+**/geometry/lattices/lattice <uid>/lower_left** (*double[]*)
+
+    The coordinates of the lower-left corner of the lattice. This dataset is
+    present only when the 'type' dataset is set to 'rectangular'.
+
+**/geometry/lattices/lattice <uid>/n_rings** (*int*)
+
+    Number of radial ring positions in the xy-plane. This dataset is present
+    only when the 'type' dataset is set to 'hexagonal'.
+
+**/geometry/lattices/lattice <uid>/n_axial** (*int*)
+
+    Number of lattice positions along the z-axis. This dataset is present only
+    when the 'type' dataset is set to 'hexagonal'.
+
+**/geometry/lattices/lattice <uid>/center** (*double[]*)
+
+    Coordinates of the center of the lattice. This dataset is present only when
+    the 'type' dataset is set to 'hexagonal'.
+
+**/n_materials** (*int*)
+
+    Number of materials in the problem.
+
+**/materials/material <uid>/index** (*int*)
+
+    Index in materials array used internally in OpenMC.
+
+**/materials/material <uid>/name** (*char[]*)
+
+    Name of the material.
+
+**/materials/material <uid>/atom_density** (*double[]*)
+
+    Total atom density of the material in atom/b-cm.
+
+**/materials/material <uid>/nuclides** (*char[][]*)
+
+    Array of nuclides present in the material, e.g., 'U-235.71c'.
+
+**/materials/material <uid>/nuclide_densities** (*double[]*)
+
+    Atom density of each nuclide.
+
+**/materials/material <uid>/sab_names** (*char[][]*)
+
+    Names of S(:math:`\alpha`,:math:`\beta`) tables assigned to the material.
+
+**/tallies/n_tallies** (*int*)
+
+    Number of tallies in the problem.
+
+**/tallies/n_meshes** (*int*)
+
+    Number of meshes in the problem.
+
+**/tallies/mesh <uid>/index** (*int*)
+
+    Index in the meshes array used internally in OpenMC.
+
+**/tallies/mesh <uid>/type** (*char[]*)
+
+    Type of the mesh. The only valid option is currently 'regular'.
+
+**/tallies/mesh <uid>/dimension** (*int[]*)
+
+    Number of mesh cells in each direction.
+
+**/tallies/mesh <uid>/lower_left** (*double[]*)
+
+    Coordinates of the lower-left corner of the mesh.
+
+**/tallies/mesh <uid>/upper_right** (*double[]*)
+
+    Coordinates of the upper-right corner of the mesh.
+
+**/tallies/mesh <uid>/width** (*double[]*)
+
+    Width of a single mesh cell in each direction.
+
+**/tallies/tally <uid>/index** (*int*)
+
+    Index in tallies array used internally in OpenMC.
+
+**/tallies/tally <uid>/name** (*char[]*)
+
+    Name of the tally.
+
+**/tallies/tally <uid>/n_filters** (*int*)
+
+    Number of filters applied to the tally.
+
+**/tallies/tally <uid>/filter <j>/type** (*char[]*)
+
+    Type of the j-th filter. Can be 'universe', 'material', 'cell', 'cellborn',
+    'surface', 'mesh', 'energy', 'energyout', or 'distribcell'.
+
+**/tallies/tally <uid>/filter <j>/offset** (*int*)
+
+    Filter offset (used for distribcell filter).
+
+**/tallies/tally <uid>/filter <j>/n_bins** (*int*)
+
+    Number of bins for the j-th filter.
+
+**/tallies/tally <uid>/filter <j>/bins** (*int[]* or *double[]*)
+
+    Value for each filter bin of this type.
+
+**/tallies/tally <uid>/nuclides** (*char[][]*)
+
+    Array of nuclides to tally. Note that if no nuclide is specified in the user
+    input, a single 'total' nuclide appears here.
+
+**/tallies/tally <uid>/n_score_bins** (*int*)
+
+    Number of scoring bins for a single nuclide. In general, this can be greater
+    than the number of user-specified scores since each score might have
+    multiple scoring bins, e.g., scatter-PN.
+
+**/tallies/tally <uid>/score_bins** (*char[][]*)
+
+    Scoring bins for the tally.

_sources/usersguide/output/track.txt

@@ -0,0 +1,30 @@
+.. _usersguide_track:
+
+=================
+Track File Format
+=================
+
+The current revision of the particle track file format is 1.
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/revision** (*int*)
+
+    Revision of the track file format. Any time a change is made in the format,
+    this integer is incremented.
+
+**/n_particles** (*int*)
+
+    Number of particles for which tracks are recorded.
+
+**/n_coords** (*int[]*)
+
+    Number of coordinates for each particle.
+
+*do i = 1, n_particles*
+
+    **/coordinates_i** (*double[][3]*)
+
+        (x,y,z) coordinates for the *i*-th particle.

_sources/usersguide/output/voxel.txt

@@ -0,0 +1,25 @@
+.. _usersguide_voxel:
+
+======================
+Voxel Plot File Format
+======================
+
+**/filetype** (*char[]*)
+
+    String indicating the type of file.
+
+**/num_voxels** (*int[3]*)
+
+    Number of voxels in the x-, y-, and z- directions.
+
+**/voxel_width** (*double[3]*)
+
+    Width of a voxel in centimeters.
+
+**/lower_left** (*double[3]*)
+
+    Cartesian coordinates of the lower-left corner of the plot.
+
+**/data** (*int[][][]*)
+
+    Data for each voxel that represents a material or cell ID.

_sources/usersguide/processing.txt

@@ -6,31 +6,34 @@ Data Processing and Visualization
 
 This section is intended to explain in detail the recommended procedures for
 carrying out common post-processing tasks with OpenMC. While several utilities
-of varying complexity are provided to help automate the process, in many cases
-it will be extremely beneficial to do some coding in Python to quickly obtain
-results. In these cases, and for many of the provided utilities, it is necessary
-for your Python installation to contain:
+of varying complexity are provided to help automate the process, the most
+powerful capabilities for post-processing derive from use of the :ref:`Python
+API <pythonapi>`. Both the provided scripts and the Python API rely on a number
+third-party Python packages, including:
 
-* [1]_ `Numpy <http://www.numpy.org/>`_
-* [1]_ `Scipy <http://www.scipy.org/>`_
-* [2]_ `h5py <http://code.google.com/p/h5py/>`_
-* [3]_ `Matplotlib <http://matplotlib.org/>`_
-* [3]_ `Silomesh <https://github.com/nhorelik/silomesh>`_
-* [3]_ `VTK <http://www.vtk.org/>`_
+* [1]_ `NumPy <http://www.numpy.org/>`_
+* [2]_ `h5py <http://www.h5py.org>`_
+* [3]_ `pandas <http://pandas.pydata.org>`_
+* [4]_ `matplotlib <http://matplotlib.org/>`_
+* [4]_ `Silomesh <https://github.com/nhorelik/silomesh>`_
+* [4]_ `VTK <http://www.vtk.org/>`_
+* [4]_ `lxml <http://lxml.de>`_
 
-Most of these are easily obtainable in Ubuntu through the package manager, or
-are easily installed with distutils.
+Most of these are can easily be installed with `pip <https://pip.pypa.io>`_
+or alternatively obtaining through a package manager.
 
-.. [1] Required for tally data extraction from statepoints with statepoint.py
-.. [2] Required only if reading HDF5 statepoint files.
-.. [3] Optional for plotting utilities
+.. [1] Required for most post-processing tasks
+.. [2] Required for reading HDF5 output files
+.. [3] Optional dependency for advanced features in Python API
+.. [4] Not used directly by the Python API, but are optional dependencies for a
+       number of scripts.
 
 ----------------------
 Geometry Visualization
 ----------------------
 
 Geometry plotting is carried out by creating a plots.xml, specifying plots, and
-running OpenMC with the -plot or -p command-line option (See
+running OpenMC with the --plot or -p command-line option (See
 :ref:`usersguide_plotting`).
 
 Plotting in 2D
@@ -128,27 +131,26 @@ capabilities of 3D voxel plots.
 Voxel plots are built the same way 2D slice plots are, by determining the cell
 or material id of a particle at the center of each voxel. In this example, the
 space covered is the cube between the points (-5,-5,-5) and (5,5,5), with voxel
-centers 10/500 = 0.02 cm apart. The binary VOXEL files that are produced do not
+centers 10/500 = 0.02 cm apart. The HDF5 voxel files that are produced do not
 specify any color - instead containing only material or cell ids (material id
 in this example) - and thus the ``background``, ``col_spec``, and ``mask``
 elements are not used. If no cell is found at a voxel center, an id of -1 is
 stored.
 
-The binary VOXEL files output by OpenMC can not be viewed directly by any
-existing viewers. In order to view them, they must be converted into a standard
-mesh format that can be viewed in ParaView, Visit, etc. This typically will
-compress the size of the file significantly. The provided utility voxel.py
-accomplishes this for SILO:
+The voxel plot data is written to an HDF5 file. The voxel file can subsequently
+be converted into a standard mesh format that can be viewed in ParaView, Visit,
+etc. This typically will compress the size of the file significantly. The
+provided utility openmc-voxel-to-silovtk accomplishes this for SILO:
 
 .. code-block:: sh
 
-    <openmc_root>/src/utils/voxel.py myplot.voxel -o output.silo
+    openmc-voxel-to-silovtk myplot.voxel -o output.silo
 
 and VTK file formats:
 
 .. code-block:: sh
 
-    <openmc_root>/src/utils/voxel.py myplot.voxel --vtk -o output.vti
+    openmc-voxel-to-silovtk myplot.voxel --vtk -o output.vti
 
 To use this utility you need either
 
@@ -156,11 +158,10 @@ To use this utility you need either
 
 or
 
-* `VTK <http://www.vtk.org/>`_ with python bindings - On Ubuntu, these are
-  easily obtained with ``sudo apt-get install python-vtk``
+* `VTK <http://www.vtk.org/>`_ with python bindings. On debian derivatives,
+  these are easily obtained with ``sudo apt-get install python-vtk``
 
-Users can process the binary into any other format if desired by following the
-example of voxel.py.  For the binary file structure, see :ref:`devguide_voxel`.
+For the HDF5 file structure, see :ref:`usersguide_voxel`.
 
 Once processed into a standard 3D file format, colors and masks can be defined
 using the stored id numbers to better explore the geometry. The process for
@@ -183,150 +184,38 @@ doing this will depend on the 3D viewer, but should be straightforward.
 Tally Visualization
 -------------------
 
-Tally results are saved in both a text file (tallies.out) as well as a binary
+Tally results are saved in both a text file (tallies.out) as well as an HDF5
 statepoint file. While the tallies.out file may be fine for simple tallies, in
-many cases the user requires more information about the tally or the run, or
-has to deal with a large number of result values (e.g. for mesh tallies).  In
-these cases, extracting data from the statepoint file via Python scripting is
-the preferred method of data analysis and visualization.
+many cases the user requires more information about the tally or the run, or has
+to deal with a large number of result values (e.g. for mesh tallies).  In these
+cases, extracting data from the statepoint file via the :ref:`pythonapi` is the
+preferred method of data analysis and visualization.
 
 Data Extraction
 ---------------
 
 A great deal of information is available in statepoint files (See
-:ref:`devguide_statepoint`), most of which is easily extracted by the provided
-utility statepoint.py. This utility provides a Python class to load statepoints
-and extract data - it is used in many of the provided plotting utilities, and
-can be used in user-created scripts to carry out manipulations of the data. To
-read tallies using this utility, make sure statepoint.py is in your PYTHONPATH,
-and then import the class, instantiate it, and call read_results:
+:ref:`usersguide_statepoint`), all of which is accessible through the Python
+API. The ``openmc.statepoint`` module (see :ref:`pythonapi_statepoint`) provides
+a class to load statepoints and access data as requested; it is used in many of
+the provided plotting utilities, OpenMC's regression test suite, and can be used
+in user-created scripts to carry out manipulations of the data.
 
-.. code-block:: python
-
-    from statepoint import StatePoint
-    sp = StatePoint('statepoint.100.binary')
-    sp.read_results()
-
-At this point the user can extract entire scores from tallies into a data
-dictionary containing numpy arrays:
-
-.. code-block:: python
-
-    tallyid = 1
-    score = 'flux'
-    data = sp.extract_results(tallyid, score)
-    means = data['means']
-    print data.keys()
-
-The results from this function contain all filter bins (all mesh points, all
-energy groups, etc.), which can be reshaped with the bin ordering also contained
-in the output dictionary. This is the best choice of output for easily
-integrating ranges of data.
-
-Alternatively the user can extract specific values for a single score/filter
-combination:
-
-.. code-block:: python
-
-    tallyid = 1
-    score = 'flux'
-    filters = [('mesh', (1, 1, 5)), ('energyin', 0)]
-    value, error = sp.get_value(tallyid, filters, score)
-
-In the future more documentation may become available here for statepoint.py and
-the data extraction functions of StatePoint objects. However, for now it is up
-to the user to explore the classes in statepoint.py to discover what data is
-available in StatePoint objects (we highly recommend interactively exploring
-with `IPython <http://ipython.org/>`_). Many examples can be found by looking
-through the other utilities that use statepoint.py, and a few common
-visualization tasks will be described here in the following sections.
+An :ref:`example IPython notebook <notebook_post_processing>` demonstrates how
+to extract data from a statepoint using the Python API.
 
 Plotting in 2D
 --------------
 
+The :ref:`IPython notebook example <notebook_post_processing>` also demonstrates
+how to plot a mesh tally in two dimensions using the Python API. Note, however,
+that there is also a script distributed with OpenMC, ``openmc-plot-mesh-tally``,
+that provides an interactive GUI to explore and plot mesh tallies for any scores
+and filter bins.
+
 .. image:: ../_images/plotmeshtally.png
    :height: 200px
 
-For simple viewing of 2D slices of a mesh plot, the utility plot_mesh_tally.py
-is provided.  This utility provides an interactive GUI to explore and plot
-mesh tallies for any scores and filter bins.  It requires statepoint.py.
-
-.. image:: ../_images/fluxplot.png
-   :height: 200px
-
-Alternatively, the user can write their own Python script to manipulate the data
-appropriately. Consider a run where the first tally contains a 105x105x20 mesh
-over a small core, with a flux score and two energyin filter bins. To explicitly
-extract the data and create a plot with gnuplot, the following script can be
-used. The script operates in several steps for clarity, and is not necessarily
-the most efficient way to extract data from large mesh tallies. This creates the
-two heatmaps in the previous figure.
-
-.. code-block:: python
-
-    #!/usr/bin/env python
-
-    import os
-
-    import statepoint
-
-    # load and parse the statepoint file
-    sp = statepoint.StatePoint('statepoint.300.binary')
-    sp.read_results()
-
-    tallyid = 0 # This is tally 1
-    score = 0   # This corresponds to flux (see tally.scores)
-
-    # get mesh dimensions
-    meshid = sp.tallies[tallyid].filters['mesh'].bins[0]
-    for i,m in enumerate(sp.meshes):
-        if m.id == meshid:
-          mesh = m
-          break
-    nx,ny,nz = mesh.dimension
-
-    # loop through mesh and extract values to python dictionaries
-    thermal = {}
-    fast = {}
-    for x in range(1,nx+1):
-        for y in range(1,ny+1):
-            for z in range(1,nz+1):
-                val,err = sp.get_value(tallyid,
-                                       [('mesh',(x,y,z)),('energyin',0)],
-                                       score)
-                thermal[(x,y,z)] = val
-                val,err = sp.get_value(tallyid,
-                                       [('mesh',(x,y,z)),('energyin',1)],
-                                       score)
-                fast[(x,y,z)] = val
-
-    # sum up the axial values and write datafile for gnuplot
-    with open('meshdata.dat','w') as fh:
-        for x in range(1,nx+1):
-            for y in range(1,ny+1):
-                thermalval = 0.
-                fastval = 0.
-                for z in range(1,nz+1):
-                  thermalval += thermal[(x,y,z)]
-                  fastval += fast[(x,y,z)]
-                fh.write("{} {} {} {}\n".format(x,y,thermalval,fastval))
-
-    # write gnuplot file
-    with open('tmp.gnuplot','w') as fh:
-      fh.write(r"""set terminal png size 1000 400
-  set output 'fluxplot.png'
-  set nokey
-  set autoscale fix
-  set multiplot layout 1,2 title "Pin Mesh Flux Tally"
-  set title "Thermal"
-  plot 'meshdata.dat' using 1:2:3 with image
-  set title "Fast"
-  plot 'meshdata.dat' using 1:2:4 with image
-  """)
-
-    # make plot
-    os.system("gnuplot < tmp.gnuplot")
-
 Plotting in 3D
 --------------
 
@@ -334,22 +223,23 @@ Plotting in 3D
    :height: 200px
 
 As with 3D plots of the geometry, meshtally data needs to be put into a standard
-format for viewing. The utility statepoint_3d.py is provided to accomplish this
-for both VTK and SILO. By default statepoint_3d.py processes a statepoint into a
-3D file with all mesh tallies and filter/score combinations,
+format for viewing. The utility ``openmc-statepoint-3d`` is provided to
+accomplish this for both VTK and SILO. By default ``openmc-statepoint-3d``
+processes a statepoint into a 3D file with all mesh tallies and filter/score
+combinations,
 
 .. code-block:: sh
 
-    <openmc_root>/src/utils/statepoint_3d.py <statepoint_file> -o output.silo
-    <openmc_root>/src/utils/statepoint_3d.py <statepoint_file> --vtk -o output.vtm
+    openmc-statepoint-3d <statepoint_file> -o output.silo
+    openmc-statepoint-3d <statepoint_file> --vtk -o output.vtm
 
 but it also provides several command-line options to selectively process only
 certain data arrays in order to keep file sizes down.
 
 .. code-block:: sh
 
-    statepoint_3d.py <statepoint_file> --tallies 2,4 --scores 4.1,4.3 -o output.silo
-    statepoint_3d.py <statepoint_file> --filters 2.energyin.1 --vtk -o output.vtm
+    openmc-statepoint-3d <statepoint_file> --tallies 2,4 --scores 4.1,4.3 -o output.silo
+    openmc-statepoint-3d <statepoint_file> --filters 2.energyin.1 --vtk -o output.vtm
 
 All available options for specifying a subset of tallies, scores, and filters
 can be listed with the ``--list`` or ``-l`` command line options.
@@ -426,13 +316,11 @@ Getting Data into MATLAB
 ------------------------
 
 There is currently no front-end utility to dump tally data to MATLAB files, but
-the process is straightforward. First extract the data using a custom Python
-script with statepoint.py, put the data into appropriately-shaped numpy arrays,
-and then use the `Scipy MATLAB IO routines
+the process is straightforward. First extract the data using the Python API via
+``openmc.statepoint`` and then use the `Scipy MATLAB IO routines
 <http://docs.scipy.org/doc/scipy/reference/tutorial/io.html>`_ to save to a MAT
-file. Note that the data contained in the output from
-``StatePoint.extract_result`` is already in a Numpy array that can be reshaped
-and dumped to MATLAB in one step.
+file. Note that all arrays that are accessible in a statepoint are already in
+NumPy arrays that can be reshaped and dumped to MATLAB in one step.
 
 ----------------------------
 Particle Track Visualization
@@ -463,15 +351,15 @@ particle numbers, respectively.  For example, to output the tracks for particles
       </track>
 
 After running OpenMC, the directory should contain a file of the form
-"track_(batch #)_(generation #)_(particle #).(binary or h5)" for each particle
-tracked.  These track files can be converted into VTK poly data files with the
-"track.py" utility.  The usage of track.py is of the form "track.py [-o OUT] IN"
-where OUT is the optional output filename and IN is one or more filenames
-describing track files.  The default output name is "track.pvtp".  A common
-usage of track.py is "track.py track*.binary" which will use the data from all
-binary track files in the directory to write a "track.pvtp" VTK output file.
-The .pvtp file can then be read and plotted by 3d visualization programs such as
-ParaView.
+"track_(batch #)_(generation #)_(particle #).h5" for each particle tracked.
+These track files can be converted into VTK poly data files with the
+``openmc-track-to-vtk`` utility.  The usage of ``openmc-track-to-vtk`` is of the
+form "openmc-track-to-vtk [-o OUT] IN" where OUT is the optional output filename
+and IN is one or more filenames describing track files.  The default output name
+is "track.pvtp".  A common usage of track.py is "openmc-track-to-vtk track*.h5"
+which will use the data from all binary track files in the directory to write a
+"track.pvtp" VTK output file.  The .pvtp file can then be read and plotted by 3d
+visualization programs such as ParaView.
 
 ----------------------
 Source Site Processing
@@ -480,43 +368,6 @@ Source Site Processing
 For eigenvalue problems, OpenMC will store information on the fission source
 sites in the statepoint file by default. For each source site, the weight,
 position, sampled direction, and sampled energy are stored. To extract this data
-from a statepoint file, the statepoint.py Python module can be used. Below is an
-example of an interactive ipython session using the statepoint.py Python module:
-
-.. code-block:: python
-
-    In [1]: import statepoint
-
-    In [2]: sp = statepoint.StatePoint('statepoint.100.h5')
-
-    In [3]: sp.read_source()
-
-    In [4]: len(sp.source)
-    Out[4]: 1000
-
-    In [5]: sp.source[0:10]
-    Out[5]:
-    [<SourceSite: xyz=[  2.21980946  -8.92686048  87.93720485] at E=0.932923263566>,
-     <SourceSite: xyz=[  2.21980946  -8.92686048  87.93720485] at E=0.349240220512>,
-     <SourceSite: xyz=[-31.21542213 -30.26762771  72.10845757] at E=3.75843584486>,
-     <SourceSite: xyz=[-31.21542213 -30.26762771  72.10845757] at E=0.80550137267>,
-     <SourceSite: xyz=[   0.18805099  -69.13376508  103.67726838] at E=1.67922461097>,
-     <SourceSite: xyz=[   0.18805099  -69.13376508  103.67726838] at E=1.16304110199>,
-     <SourceSite: xyz=[ -50.42189115   -9.96571672  123.34077905] at E=0.710937974074>,
-     <SourceSite: xyz=[ -32.80427668  -15.49316628  125.26301151] at E=1.61907104162>,
-     <SourceSite: xyz=[  53.20376026  -15.38643708  120.58071044] at E=3.33962024907>,
-     <SourceSite: xyz=[  53.20376026  -15.38643708  120.58071044] at E=1.90185680329>]
-
-    In [6]: site = sp.source[0]
-
-    In [7]: site.weight
-    Out[7]: 1.0
-
-    In [8]: site.xyz
-    Out[8]: array([  2.21980946,  -8.92686048,  87.93720485])
-
-    In [9]: site.uvw
-    Out[9]: array([ 0.06740523,  0.50612814,  0.85982024])
-
-    In [10]: site.E
-    Out[10]: 0.93292326356564159
+from a statepoint file, the ``openmc.statepoint`` module can be used. An
+:ref:`example IPython notebook <notebook_post_processing>` demontrates how to
+analyze and plot source information.

_sources/usersguide/troubleshoot.txt

@@ -31,21 +31,6 @@ f951: error: unrecognized command line option "-fbacktrace"
 You are probably using a version of the gfortran compiler that is too
 old. Download and install the latest version of gfortran_.
 
-
-make[1]: ifort: Command not found
-*********************************
-
-You tried compiling with the Intel Fortran compiler and it was not found on your
-:envvar:`PATH`. If you have the Intel compiler installed, make sure the shell
-can locate it (this can be tested with :program:`which ifort`).
-
-make[1]: pgf90: Command not found
-*********************************
-
-You tried compiling with the PGI Fortran compiler and it was not found on your
-:envvar:`PATH`. If you have the PGI compiler installed, make sure the shell can
-locate it (this can be tested with :program:`which pgf90`).
-
 -------------------------
 Problems with Simulations
 -------------------------
@@ -56,13 +41,13 @@ Segmentation Fault
 A segmentation fault occurs when the program tries to access a variable in
 memory that was outside the memory allocated for the program. The best way to
 debug a segmentation fault is to re-compile OpenMC with debug options turned
-on. First go to your ``openmc/src`` directory where OpenMC was compiled and type
-the following commands:
+on. Create a new build directory and type the following commands:
 
 .. code-block:: sh
 
-    make distclean
-    make DEBUG=yes
+    mkdir build-debug && cd build-debug
+    cmake -Ddebug=on /path/to/openmc
+    make
 
 Now when you re-run your problem, it should report exactly where the program
 failed. If after reading the debug output, you are still unsure why the program

_static/searchtools.js

@@ -10,6 +10,7 @@
  */
 
 
+/* Non-minified version JS is _stemmer.js if file is provided */ 
 /**
  * Porter Stemmer
  */
@@ -373,8 +374,7 @@ var Search = {
     }
 
     // lookup as search terms in fulltext
-    results = results.concat(this.performTermsSearch(searchterms, excluded, terms, Scorer.term))
-                     .concat(this.performTermsSearch(searchterms, excluded, titleterms, Scorer.title));
+    results = results.concat(this.performTermsSearch(searchterms, excluded, terms, titleterms));
 
     // let the scorer override scores with a custom scoring function
     if (Scorer.score) {
@@ -538,23 +538,47 @@ var Search = {
   /**
    * search for full-text terms in the index
    */
-  performTermsSearch : function(searchterms, excluded, terms, score) {
+  performTermsSearch : function(searchterms, excluded, terms, titleterms) {
     var filenames = this._index.filenames;
     var titles = this._index.titles;
 
-    var i, j, file, files;
+    var i, j, file;
     var fileMap = {};
+    var scoreMap = {};
     var results = [];
 
     // perform the search on the required terms
     for (i = 0; i < searchterms.length; i++) {
       var word = searchterms[i];
+      var files = [];
+      var _o = [
+        {files: terms[word], score: Scorer.term},
+        {files: titleterms[word], score: Scorer.title}
+      ];
+
       // no match but word was a required one
-      if ((files = terms[word]) === undefined)
+      if ($u.every(_o, function(o){return o.files === undefined;})) {
         break;
-      if (files.length === undefined) {
-        files = [files];
       }
+      // found search word in contents
+      $u.each(_o, function(o) {
+        var _files = o.files;
+        if (_files === undefined)
+          return
+
+        if (_files.length === undefined)
+          _files = [_files];
+        files = files.concat(_files);
+
+        // set score for the word in each file to Scorer.term
+        for (j = 0; j < _files.length; j++) {
+          file = _files[j];
+          if (!(file in scoreMap))
+            scoreMap[file] = {}
+          scoreMap[file][word] = o.score;
+        }
+      });
+
       // create the mapping
       for (j = 0; j < files.length; j++) {
         file = files[j];
@@ -576,7 +600,9 @@ var Search = {
       // ensure that none of the excluded terms is in the search result
       for (i = 0; i < excluded.length; i++) {
         if (terms[excluded[i]] == file ||
-          $u.contains(terms[excluded[i]] || [], file)) {
+            titleterms[excluded[i]] == file ||
+            $u.contains(terms[excluded[i]] || [], file) ||
+            $u.contains(titleterms[excluded[i]] || [], file)) {
           valid = false;
           break;
         }
@@ -584,6 +610,9 @@ var Search = {
 
       // if we have still a valid result we can add it to the result list
       if (valid) {
+        // select one (max) score for the file.
+        // for better ranking, we should calculate ranking by using words statistics like basic tf-idf...
+        var score = $u.max($u.map(fileMap[file], function(w){return scoreMap[file][w]}));
         results.push([filenames[file], titles[file], '', null, score]);
       }
     }

developers.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    './',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -76,7 +76,7 @@
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

devguide/docbuild.html

@@ -6,7 +6,7 @@
   <head>
     <meta http-equiv="Content-Type" content="text/html; charset=utf-8" />
     
-    <title>7. Building Sphinx Documentation &mdash; OpenMC Documentation</title>
+    <title>5. Building Sphinx Documentation &mdash; OpenMC Documentation</title>
     
     <link rel="stylesheet" href="../_static/haiku.css" type="text/css" />
     <link rel="stylesheet" href="../_static/pygments.css" type="text/css" />
@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -26,7 +26,7 @@
     <link rel="top" title="OpenMC Documentation" href="../index.html" />
     <link rel="up" title="Developer’s Guide" href="index.html" />
     <link rel="next" title="Python API" href="../pythonapi/index.html" />
-    <link rel="prev" title="6. Voxel Plot Binary File Specifications" href="voxel.html" /> 
+    <link rel="prev" title="4. XML Input Parsing" href="xml-parsing.html" /> 
   </head>
   <body role="document">
       <div class="header" role="banner">
@@ -37,7 +37,7 @@
       <div class="topnav" role="navigation" aria-label="top navigation">
       
         <p>
-        «&#160;&#160;<a href="voxel.html">6. Voxel Plot Binary File Specifications</a>
+        «&#160;&#160;<a href="xml-parsing.html">4. XML Input Parsing</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -49,7 +49,7 @@
         
         
   <div class="section" id="building-sphinx-documentation">
-<span id="devguide-docbuild"></span><h1>7. Building Sphinx Documentation<a class="headerlink" href="#building-sphinx-documentation" title="Permalink to this headline">¶</a></h1>
+<span id="devguide-docbuild"></span><h1>5. Building Sphinx Documentation<a class="headerlink" href="#building-sphinx-documentation" title="Permalink to this headline">¶</a></h1>
 <p>In order to build the documentation in the <code class="docutils literal"><span class="pre">docs</span></code> directory, you will need to
 have the <a class="reference external" href="http://sphinx-doc.org">Sphinx</a> third-party Python package. The easiest way to install Sphinx
 is via pip:</p>
@@ -67,7 +67,7 @@ from the git repository as such:</p>
 </pre></div>
 </div>
 <div class="section" id="building-documentation-as-a-webpage">
-<h2>7.1. Building Documentation as a Webpage<a class="headerlink" href="#building-documentation-as-a-webpage" title="Permalink to this headline">¶</a></h2>
+<h2>5.1. Building Documentation as a Webpage<a class="headerlink" href="#building-documentation-as-a-webpage" title="Permalink to this headline">¶</a></h2>
 <p>To build the documentation as a webpage (what appears at
 <a class="reference external" href="http://mit-crpg.github.io/openmc">http://mit-crpg.github.io/openmc</a>), simply go to the <code class="docutils literal"><span class="pre">docs</span></code> directory and run:</p>
 <div class="highlight-sh"><div class="highlight"><pre>make html
@@ -75,7 +75,7 @@ from the git repository as such:</p>
 </div>
 </div>
 <div class="section" id="building-documentation-as-a-pdf">
-<h2>7.2. Building Documentation as a PDF<a class="headerlink" href="#building-documentation-as-a-pdf" title="Permalink to this headline">¶</a></h2>
+<h2>5.2. Building Documentation as a PDF<a class="headerlink" href="#building-documentation-as-a-pdf" title="Permalink to this headline">¶</a></h2>
 <p>To build PDF documentation, you will need to have a LaTeX distribution installed
 on your computer as well as <a class="reference external" href="https://inkscape.org">Inkscape</a>, which is used to convert .svg files to
 .pdf files. Inkscape can be installed in a Debian-derivative with:</p>
@@ -95,7 +95,7 @@ run:</p>
       <div class="bottomnav" role="navigation" aria-label="bottom navigation">
       
         <p>
-        «&#160;&#160;<a href="voxel.html">6. Voxel Plot Binary File Specifications</a>
+        «&#160;&#160;<a href="xml-parsing.html">4. XML Input Parsing</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -107,7 +107,7 @@ run:</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

devguide/index.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -25,7 +25,7 @@
     <script type="text/javascript" src="../_static/doctools.js"></script>
     <link rel="top" title="OpenMC Documentation" href="../index.html" />
     <link rel="next" title="1. Data Structures" href="structures.html" />
-    <link rel="prev" title="5. Troubleshooting" href="../usersguide/troubleshoot.html" /> 
+    <link rel="prev" title="6. Troubleshooting" href="../usersguide/troubleshoot.html" /> 
   </head>
   <body role="document">
       <div class="header" role="banner">
@@ -36,7 +36,7 @@
       <div class="topnav" role="navigation" aria-label="top navigation">
       
         <p>
-        «&#160;&#160;<a href="../usersguide/troubleshoot.html">5. Troubleshooting</a>
+        «&#160;&#160;<a href="../usersguide/troubleshoot.html">6. Troubleshooting</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -95,11 +95,9 @@ as debugging.</p>
 <li class="toctree-l2"><a class="reference internal" href="xml-parsing.html#editing-fox-on-personal-fork">4.2. Editing FoX on Personal Fork</a></li>
 </ul>
 </li>
-<li class="toctree-l1"><a class="reference internal" href="statepoint.html">5. State Point Binary File Specifications</a></li>
-<li class="toctree-l1"><a class="reference internal" href="voxel.html">6. Voxel Plot Binary File Specifications</a></li>
-<li class="toctree-l1"><a class="reference internal" href="docbuild.html">7. Building Sphinx Documentation</a><ul>
-<li class="toctree-l2"><a class="reference internal" href="docbuild.html#building-documentation-as-a-webpage">7.1. Building Documentation as a Webpage</a></li>
-<li class="toctree-l2"><a class="reference internal" href="docbuild.html#building-documentation-as-a-pdf">7.2. Building Documentation as a PDF</a></li>
+<li class="toctree-l1"><a class="reference internal" href="docbuild.html">5. Building Sphinx Documentation</a><ul>
+<li class="toctree-l2"><a class="reference internal" href="docbuild.html#building-documentation-as-a-webpage">5.1. Building Documentation as a Webpage</a></li>
+<li class="toctree-l2"><a class="reference internal" href="docbuild.html#building-documentation-as-a-pdf">5.2. Building Documentation as a PDF</a></li>
 </ul>
 </li>
 </ul>
@@ -111,7 +109,7 @@ as debugging.</p>
       <div class="bottomnav" role="navigation" aria-label="bottom navigation">
       
         <p>
-        «&#160;&#160;<a href="../usersguide/troubleshoot.html">5. Troubleshooting</a>
+        «&#160;&#160;<a href="../usersguide/troubleshoot.html">6. Troubleshooting</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -123,7 +121,7 @@ as debugging.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

devguide/statepoint.html

@@ -1,328 +0,0 @@
-<!DOCTYPE html PUBLIC "-//W3C//DTD XHTML 1.0 Transitional//EN"
-  "http://www.w3.org/TR/xhtml1/DTD/xhtml1-transitional.dtd">
-
-
-<html xmlns="http://www.w3.org/1999/xhtml">
-  <head>
-    <meta http-equiv="Content-Type" content="text/html; charset=utf-8" />
-    
-    <title>5. State Point Binary File Specifications &mdash; OpenMC Documentation</title>
-    
-    <link rel="stylesheet" href="../_static/haiku.css" type="text/css" />
-    <link rel="stylesheet" href="../_static/pygments.css" type="text/css" />
-    
-    <script type="text/javascript">
-      var DOCUMENTATION_OPTIONS = {
-        URL_ROOT:    '../',
-        VERSION:     '0.7.0',
-        COLLAPSE_INDEX: false,
-        FILE_SUFFIX: '.html',
-        HAS_SOURCE:  true
-      };
-    </script>
-    <script type="text/javascript" src="../_static/jquery.js"></script>
-    <script type="text/javascript" src="../_static/underscore.js"></script>
-    <script type="text/javascript" src="../_static/doctools.js"></script>
-    <link rel="top" title="OpenMC Documentation" href="../index.html" />
-    <link rel="up" title="Developer’s Guide" href="index.html" />
-    <link rel="next" title="6. Voxel Plot Binary File Specifications" href="voxel.html" />
-    <link rel="prev" title="4. XML Input Parsing" href="xml-parsing.html" /> 
-  </head>
-  <body role="document">
-      <div class="header" role="banner">
-        <a href="../index.html">
-          <img class="logo" src="../_static/openmc.png" alt="Logo"/>
-        </a>
-      </div>
-      <div class="topnav" role="navigation" aria-label="top navigation">
-      
-        <p>
-        «&#160;&#160;<a href="xml-parsing.html">4. XML Input Parsing</a>
-        &#160;&#160;::&#160;&#160;
-        <a class="uplink" href="../index.html">Contents</a>
-        &#160;&#160;::&#160;&#160;
-        <a href="voxel.html">6. Voxel Plot Binary File Specifications</a>&#160;&#160;»
-        </p>
-
-      </div>
-      <div class="content">
-        
-        
-  <div class="section" id="state-point-binary-file-specifications">
-<span id="devguide-statepoint"></span><h1>5. State Point Binary File Specifications<a class="headerlink" href="#state-point-binary-file-specifications" title="Permalink to this headline">¶</a></h1>
-<p>The current revision of the statepoint binary file is 13.</p>
-<p><strong>integer(4) FILETYPE_STATEPOINT</strong></p>
-<blockquote>
-<div>Flags whether this file is a statepoint file or a particle restart file.</div></blockquote>
-<p><strong>integer(4) REVISION_STATEPOINT</strong></p>
-<blockquote>
-<div>Revision of the binary state point file. Any time a change is made in the
-format of the state-point file, this integer is incremented.</div></blockquote>
-<p><strong>integer(4) VERSION_MAJOR</strong></p>
-<blockquote>
-<div>Major version number for OpenMC</div></blockquote>
-<p><strong>integer(4) VERSION_MINOR</strong></p>
-<blockquote>
-<div>Minor version number for OpenMC</div></blockquote>
-<p><strong>integer(4) VERSION_RELEASE</strong></p>
-<blockquote>
-<div>Release version number for OpenMC</div></blockquote>
-<p><strong>character(19) time_stamp</strong></p>
-<blockquote>
-<div>Date and time the state point was written.</div></blockquote>
-<p><strong>character(255) path</strong></p>
-<blockquote>
-<div>Absolute path to directory containing input files.</div></blockquote>
-<p><strong>integer(8) seed</strong></p>
-<blockquote>
-<div>Pseudo-random number generator seed.</div></blockquote>
-<p><strong>integer(4) run_mode</strong></p>
-<blockquote>
-<div>run mode used. The modes are described in constants.F90.</div></blockquote>
-<p><strong>integer(8) n_particles</strong></p>
-<blockquote>
-<div>Number of particles used per generation.</div></blockquote>
-<p><strong>integer(4) current_batch</strong></p>
-<blockquote>
-<div>The number of batches already simulated.</div></blockquote>
-<p>if (run_mode == MODE_EIGENVALUE)</p>
-<blockquote>
-<div><p><strong>integer(4) n_inactive</strong></p>
-<blockquote>
-<div>Number of inactive batches</div></blockquote>
-<p><strong>integer(4) gen_per_batch</strong></p>
-<blockquote>
-<div>Number of generations per batch for criticality calculations</div></blockquote>
-<p><em>do i = 1, current_batch * gen_per_batch</em></p>
-<blockquote>
-<div><p><strong>real(8) k_generation(i)</strong></p>
-<blockquote>
-<div>k-effective for the i-th total generation</div></blockquote>
-</div></blockquote>
-<p><em>do i = 1, current_batch * gen_per_batch</em></p>
-<blockquote>
-<div><p><strong>real(8) entropy(i)</strong></p>
-<blockquote>
-<div>Shannon entropy for the i-th total generation</div></blockquote>
-</div></blockquote>
-<p><strong>real(8) k_col_abs</strong></p>
-<blockquote>
-<div>Sum of product of collision/absorption estimates of k-effective</div></blockquote>
-<p><strong>real(8) k_col_tra</strong></p>
-<blockquote>
-<div>Sum of product of collision/track-length estimates of k-effective</div></blockquote>
-<p><strong>real(8) k_abs_tra</strong></p>
-<blockquote>
-<div>Sum of product of absorption/track-length estimates of k-effective</div></blockquote>
-<p><strong>real(8) k_combined(2)</strong></p>
-<blockquote>
-<div>Mean and standard deviation of a combined estimate of k-effective</div></blockquote>
-<p><strong>integer(4) cmfd_on</strong></p>
-<blockquote>
-<div>Flag that cmfd is on</div></blockquote>
-<p>if (cmfd_on)</p>
-<blockquote>
-<div><p><strong>integer(4) cmfd % indices</strong></p>
-<blockquote>
-<div>Indices for cmfd mesh (i,j,k,g)</div></blockquote>
-<p><strong>real(8) cmfd % k_cmfd(1:current_batch)</strong></p>
-<blockquote>
-<div>CMFD eigenvalues</div></blockquote>
-<p><strong>real(8) cmfd % src(1:G,1:I,1:J,1:K)</strong></p>
-<blockquote>
-<div>CMFD fission source</div></blockquote>
-<p><strong>real(8) cmfd % entropy(1:current_batch)</strong></p>
-<blockquote>
-<div>CMFD estimate of Shannon entropy</div></blockquote>
-<p><strong>real(8) cmfd % balance(1:current_batch)</strong></p>
-<blockquote>
-<div>RMS of the residual neutron balance equation on CMFD mesh</div></blockquote>
-<p><strong>real(8) cmfd % dom(1:current_batch)</strong></p>
-<blockquote>
-<div>CMFD estimate of dominance ratio</div></blockquote>
-<p><strong>real(8) cmfd % scr_cmp(1:current_batch)</strong></p>
-<blockquote>
-<div>RMS comparison of difference between OpenMC and CMFD fission source</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) n_meshes</strong></p>
-<blockquote>
-<div>Number of meshes in tallies.xml file</div></blockquote>
-<p><em>do i = 1, n_meshes</em></p>
-<blockquote>
-<div><p><strong>integer(4) meshes(i) % id</strong></p>
-<blockquote>
-<div>Unique ID of mesh.</div></blockquote>
-<p><strong>integer(4) meshes(i) % type</strong></p>
-<blockquote>
-<div>Type of mesh.</div></blockquote>
-<p><strong>integer(4) meshes(i) % n_dimension</strong></p>
-<blockquote>
-<div>Number of dimensions for mesh (2 or 3).</div></blockquote>
-<p><strong>integer(4) meshes(i) % dimension(:)</strong></p>
-<blockquote>
-<div>Number of mesh cells in each dimension.</div></blockquote>
-<p><strong>real(8) meshes(i) % lower_left(:)</strong></p>
-<blockquote>
-<div>Coordinates of lower-left corner of mesh.</div></blockquote>
-<p><strong>real(8) meshes(i) % upper_right(:)</strong></p>
-<blockquote>
-<div>Coordinates of upper-right corner of mesh.</div></blockquote>
-<p><strong>real(8) meshes(i) % width(:)</strong></p>
-<blockquote>
-<div>Width of each mesh cell in each dimension.</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) n_tallies</strong></p>
-<p><em>do i = 1, n_tallies</em></p>
-<blockquote>
-<div><p><strong>integer(4) tallies(i) % id</strong></p>
-<blockquote>
-<div>Unique ID of tally.</div></blockquote>
-<p><strong>integer(4) tallies(i) % n_realizations</strong></p>
-<blockquote>
-<div>Number of realizations for the i-th tally.</div></blockquote>
-<p><strong>integer(4) size(tallies(i) % scores, 1)</strong></p>
-<blockquote>
-<div>Total number of score bins for the i-th tally</div></blockquote>
-<p><strong>integer(4) size(tallies(i) % scores, 2)</strong></p>
-<blockquote>
-<div>Total number of filter bins for the i-th tally</div></blockquote>
-<p><strong>integer(4) tallies(i) % n_filters</strong></p>
-<p><em>do j = 1, tallies(i) % n_filters</em></p>
-<blockquote>
-<div><p><strong>integer(4) tallies(i) % filter(j) % type</strong></p>
-<blockquote>
-<div>Type of tally filter.</div></blockquote>
-<p><strong>integer(4) tallies(i) % filter(j) % n_bins</strong></p>
-<blockquote>
-<div>Number of bins for filter.</div></blockquote>
-<p><strong>integer(4)/real(8) tallies(i) % filter(j) % bins(:)</strong></p>
-<blockquote>
-<div>Value for each filter bin of this type.</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) tallies(i) % n_nuclide_bins</strong></p>
-<blockquote>
-<div>Number of nuclide bins. If none are specified, this is just one.</div></blockquote>
-<p><em>do j = 1, tallies(i) % n_nuclide_bins</em></p>
-<blockquote>
-<div><p><strong>integer(4) tallies(i) % nuclide_bins(j)</strong></p>
-<blockquote>
-<div>Values of specified nuclide bins</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) tallies(i) % n_score_bins</strong></p>
-<blockquote>
-<div>Number of scoring bins.</div></blockquote>
-<p><em>do j = 1, tallies(i) % n_score_bins</em></p>
-<blockquote>
-<div><p><strong>integer(4) tallies(i) % score_bins(j)</strong></p>
-<blockquote>
-<div>Values of specified scoring bins (e.g. SCORE_FLUX).</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) tallies(i) % n_score_bins</strong></p>
-<blockquote>
-<div>Number of scoring bins without accounting for those added by
-the scatter-pn command.</div></blockquote>
-<p><em>do j = 1, tallies(i) % n_user_score_bins</em></p>
-<blockquote>
-<div><p><strong>character(8) tallies(i) % moment_order(j)</strong></p>
-<blockquote>
-<div>Tallying moment order for Legendre and spherical
-harmonic tally expansions (<em>e.g.</em>, &#8216;P2&#8217;, &#8216;Y1,2&#8217;, etc.).</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) source_present</strong></p>
-<blockquote>
-<div>Flag indicated if source bank is present in the file</div></blockquote>
-<p><strong>integer(4) n_realizations</strong></p>
-<blockquote>
-<div>Number of realizations for global tallies.</div></blockquote>
-<p><strong>integer(4) N_GLOBAL_TALLIES</strong></p>
-<blockquote>
-<div>Number of global tally scores</div></blockquote>
-<p><em>do i = 1, N_GLOBAL_TALLIES</em></p>
-<blockquote>
-<div><p><strong>real(8) global_tallies(i) % sum</strong></p>
-<blockquote>
-<div>Accumulated sum for the i-th global tally</div></blockquote>
-<p><strong>real(8) global_tallies(i) % sum_sq</strong></p>
-<blockquote>
-<div>Accumulated sum of squares for the i-th global tally</div></blockquote>
-</div></blockquote>
-<p><strong>integer(4) tallies_on</strong></p>
-<blockquote>
-<div>Flag indicated if tallies are present in the file.</div></blockquote>
-<p>if (tallies_on &gt; 0)</p>
-<blockquote>
-<div><p><em>do i = 1, n_tallies</em></p>
-<blockquote>
-<div><p><em>do k = 1, size(tallies(i) % scores, 2)</em></p>
-<blockquote>
-<div><p><em>do j = 1, size(tallies(i) % scores, 1)</em></p>
-<blockquote>
-<div><p><strong>real(8) tallies(i) % scores(j,k) % sum</strong></p>
-<blockquote>
-<div>Accumulated sum for the j-th score and k-th filter of the
-i-th tally</div></blockquote>
-<p><strong>real(8) tallies(i) % scores(j,k) % sum_sq</strong></p>
-<blockquote>
-<div>Accumulated sum of squares for the j-th score and k-th
-filter of the i-th tally</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-<p>if (run_mode == MODE_EIGENVALUE and source_present)</p>
-<blockquote>
-<div><p><em>do i = 1, n_particles</em></p>
-<blockquote>
-<div><p><strong>real(8) source_bank(i) % wgt</strong></p>
-<blockquote>
-<div>Weight of the i-th source particle</div></blockquote>
-<p><strong>real(8) source_bank(i) % xyz(1:3)</strong></p>
-<blockquote>
-<div>Coordinates of the i-th source particle.</div></blockquote>
-<p><strong>real(8) source_bank(i) % uvw(1:3)</strong></p>
-<blockquote>
-<div>Direction of the i-th source particle</div></blockquote>
-<p><strong>real(8) source_bank(i) % E</strong></p>
-<blockquote>
-<div>Energy of the i-th source particle.</div></blockquote>
-</div></blockquote>
-</div></blockquote>
-</div>
-
-
-      </div>
-      <div class="bottomnav" role="navigation" aria-label="bottom navigation">
-      
-        <p>
-        «&#160;&#160;<a href="xml-parsing.html">4. XML Input Parsing</a>
-        &#160;&#160;::&#160;&#160;
-        <a class="uplink" href="../index.html">Contents</a>
-        &#160;&#160;::&#160;&#160;
-        <a href="voxel.html">6. Voxel Plot Binary File Specifications</a>&#160;&#160;»
-        </p>
-
-      </div>
-
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-    <div class="footer" role="contentinfo">
-        &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
-    </div>
-<script type="text/javascript">
-
-  var _gaq = _gaq || [];
-  _gaq.push(['_setAccount', 'UA-30411614-1']);
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-</html>

devguide/structures.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -185,7 +185,7 @@ module</a> is of type MaterialMacroXS.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

devguide/styleguide.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -196,7 +196,7 @@ dependencies rather than required dependencies.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

devguide/voxel.html

@@ -1,132 +0,0 @@
-<!DOCTYPE html PUBLIC "-//W3C//DTD XHTML 1.0 Transitional//EN"
-  "http://www.w3.org/TR/xhtml1/DTD/xhtml1-transitional.dtd">
-
-
-<html xmlns="http://www.w3.org/1999/xhtml">
-  <head>
-    <meta http-equiv="Content-Type" content="text/html; charset=utf-8" />
-    
-    <title>6. Voxel Plot Binary File Specifications &mdash; OpenMC Documentation</title>
-    
-    <link rel="stylesheet" href="../_static/haiku.css" type="text/css" />
-    <link rel="stylesheet" href="../_static/pygments.css" type="text/css" />
-    
-    <script type="text/javascript">
-      var DOCUMENTATION_OPTIONS = {
-        URL_ROOT:    '../',
-        VERSION:     '0.7.0',
-        COLLAPSE_INDEX: false,
-        FILE_SUFFIX: '.html',
-        HAS_SOURCE:  true
-      };
-    </script>
-    <script type="text/javascript" src="../_static/jquery.js"></script>
-    <script type="text/javascript" src="../_static/underscore.js"></script>
-    <script type="text/javascript" src="../_static/doctools.js"></script>
-    <link rel="top" title="OpenMC Documentation" href="../index.html" />
-    <link rel="up" title="Developer’s Guide" href="index.html" />
-    <link rel="next" title="7. Building Sphinx Documentation" href="docbuild.html" />
-    <link rel="prev" title="5. State Point Binary File Specifications" href="statepoint.html" /> 
-  </head>
-  <body role="document">
-      <div class="header" role="banner">
-        <a href="../index.html">
-          <img class="logo" src="../_static/openmc.png" alt="Logo"/>
-        </a>
-      </div>
-      <div class="topnav" role="navigation" aria-label="top navigation">
-      
-        <p>
-        «&#160;&#160;<a href="statepoint.html">5. State Point Binary File Specifications</a>
-        &#160;&#160;::&#160;&#160;
-        <a class="uplink" href="../index.html">Contents</a>
-        &#160;&#160;::&#160;&#160;
-        <a href="docbuild.html">7. Building Sphinx Documentation</a>&#160;&#160;»
-        </p>
-
-      </div>
-      <div class="content">
-        
-        
-  <div class="section" id="voxel-plot-binary-file-specifications">
-<span id="devguide-voxel"></span><h1>6. Voxel Plot Binary File Specifications<a class="headerlink" href="#voxel-plot-binary-file-specifications" title="Permalink to this headline">¶</a></h1>
-<p>The current revision of the voxel plot binary file is 1.</p>
-<p><strong>integer(4) n_voxels_x</strong></p>
-<blockquote>
-<div>Number of voxels in the x direction</div></blockquote>
-<p><strong>integer(4) n_voxels_y</strong></p>
-<blockquote>
-<div>Number of voxels in the y direction</div></blockquote>
-<p><strong>integer(4) n_voxels_z</strong></p>
-<blockquote>
-<div>Number of voxels in the z direction</div></blockquote>
-<p><strong>real(8) width_voxel_x</strong></p>
-<blockquote>
-<div>Width of voxels in the x direction</div></blockquote>
-<p><strong>real(8) width_voxel_y</strong></p>
-<blockquote>
-<div>Width of voxels in the y direction</div></blockquote>
-<p><strong>real(8) width_voxel_z</strong></p>
-<blockquote>
-<div>Width of voxels in the z direction</div></blockquote>
-<p><strong>real(8) lower_left_x</strong></p>
-<blockquote>
-<div>Lower left x point of the voxel grid</div></blockquote>
-<p><strong>real(8) lower_left_y</strong></p>
-<blockquote>
-<div>Lower left y point of the voxel grid</div></blockquote>
-<p><strong>real(8) lower_left_z</strong></p>
-<blockquote>
-<div>Lower left z point of the voxel grid</div></blockquote>
-<dl class="docutils">
-<dt><em>do x = 1, n_voxels_x</em></dt>
-<dd><dl class="first last docutils">
-<dt><em>do y = 1, n_voxels_y</em></dt>
-<dd><p class="first"><em>do z = 1, n_voxels_z</em></p>
-<blockquote class="last">
-<div><p><strong>integer(4) id</strong></p>
-<blockquote>
-<div>Cell or material id number at this voxel center. Set to -1 when
-cell not_found.</div></blockquote>
-</div></blockquote>
-</dd>
-</dl>
-</dd>
-</dl>
-</div>
-
-
-      </div>
-      <div class="bottomnav" role="navigation" aria-label="bottom navigation">
-      
-        <p>
-        «&#160;&#160;<a href="statepoint.html">5. State Point Binary File Specifications</a>
-        &#160;&#160;::&#160;&#160;
-        <a class="uplink" href="../index.html">Contents</a>
-        &#160;&#160;::&#160;&#160;
-        <a href="docbuild.html">7. Building Sphinx Documentation</a>&#160;&#160;»
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-
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-    <div class="footer" role="contentinfo">
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devguide/workflow.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -284,7 +284,7 @@ from your private repository into a public fork.</p>
 
     <div class="footer" role="contentinfo">
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 <script type="text/javascript">
 

devguide/xml-parsing.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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         HAS_SOURCE:  true
@@ -25,7 +25,7 @@
     <script type="text/javascript" src="../_static/doctools.js"></script>
     <link rel="top" title="OpenMC Documentation" href="../index.html" />
     <link rel="up" title="Developer’s Guide" href="index.html" />
-    <link rel="next" title="5. State Point Binary File Specifications" href="statepoint.html" />
+    <link rel="next" title="5. Building Sphinx Documentation" href="docbuild.html" />
     <link rel="prev" title="3. Development Workflow" href="workflow.html" /> 
   </head>
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@@ -41,7 +41,7 @@
         &#160;&#160;::&#160;&#160;
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         &#160;&#160;::&#160;&#160;
-        <a href="statepoint.html">5. State Point Binary File Specifications</a>&#160;&#160;»
+        <a href="docbuild.html">5. Building Sphinx Documentation</a>&#160;&#160;»
         </p>
 
       </div>
@@ -155,7 +155,7 @@ were not committed.</p>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
-        <a href="statepoint.html">5. State Point Binary File Specifications</a>&#160;&#160;»
+        <a href="docbuild.html">5. Building Sphinx Documentation</a>&#160;&#160;»
         </p>
 
       </div>
@@ -163,7 +163,7 @@ were not committed.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

index.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    './',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -62,7 +62,7 @@ free to send a message to the User&#8217;s Group <a class="reference external" h
 <div class="toctree-wrapper compound">
 <ul>
 <li class="toctree-l1"><a class="reference internal" href="quickinstall.html">Quick Install Guide</a></li>
-<li class="toctree-l1"><a class="reference internal" href="releasenotes.html">Release Notes for OpenMC 0.7.0</a></li>
+<li class="toctree-l1"><a class="reference internal" href="releasenotes.html">Release Notes for OpenMC 0.7.1</a></li>
 <li class="toctree-l1"><a class="reference internal" href="methods/index.html">Theory and Methodology</a></li>
 <li class="toctree-l1"><a class="reference internal" href="usersguide/index.html">User&#8217;s Guide</a></li>
 <li class="toctree-l1"><a class="reference internal" href="devguide/index.html">Developer&#8217;s Guide</a></li>
@@ -89,7 +89,7 @@ free to send a message to the User&#8217;s Group <a class="reference external" h
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

license.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    './',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -83,7 +83,7 @@ CONNECTION WITH THE SOFTWARE OR THE USE OR OTHER DEALINGS IN THE SOFTWARE.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/cmfd.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -1043,7 +1043,7 @@ by the user. Once all diffusion parameters are calculated, CMFD matrices are
 formed where energy groups are the inner most iteration index. In OpenMC,
 compressed row storage sparse matrices are used due to the sparsity of CMFD
 operators. An example of this sparsity is shown for the 3-D BEAVRS model in
-figures <a href="#fig-loss">1</a> and <a href="#fig-prod">2</a> <a class="reference internal" href="#beavrs" id="id3">[BEAVRS]</a>. These matrices represent
+figures <a href="#fig-loss">12</a> and <a href="#fig-prod">13</a> <a class="reference internal" href="#beavrs" id="id3">[BEAVRS]</a>. These matrices represent
 an assembly radial mesh, 24 cell mesh in the axial direction and two energy
 groups. The loss matrix is 99.92% sparse and the production matrix is 99.99%
 sparse. Although the loss matrix looks like it is tridiagonal, it is really a
@@ -1101,11 +1101,11 @@ no fission neutrons appear with energies in the thermal group.</p>
 </table>
 <div class="figure" id="id12">
 <span id="fig-loss"></span><a class="reference internal image-reference" href="../_images/loss.png"><img alt="../_images/loss.png" src="../_images/loss.png" style="width: 500.0px; height: 500.0px;" /></a>
-<p class="caption"><span class="caption-text">Figure 1: Sparsity of Neutron Loss Operator</span></p>
+<p class="caption"><span class="caption-text">Figure 12: Sparsity of Neutron Loss Operator</span></p>
 </div>
 <div class="figure" id="id13">
 <span id="fig-prod"></span><a class="reference internal image-reference" href="../_images/prod.png"><img alt="../_images/prod.png" src="../_images/prod.png" style="width: 500.0px; height: 500.0px;" /></a>
-<p class="caption"><span class="caption-text">Figure 2: Sparsity of Neutron Production Operator</span></p>
+<p class="caption"><span class="caption-text">Figure 13: Sparsity of Neutron Production Operator</span></p>
 </div>
 <p>To solve the eigenvalue problem with these matrices, different source iteration
 and linear solvers can be used. The most common source iteration solver used is
@@ -1230,7 +1230,7 @@ CASMO-4E</em>. In Proceedings of PHYSOR 2002, Seoul, Korea, October 7 - 10, 2002
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/cross_sections.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
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         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -129,7 +129,7 @@ Carlo codes,&#8221; LA-UR-14-24530, Los Alamos National Laboratory (2014).</td><
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/eigenvalue.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -109,7 +109,7 @@ convergence, we would either need to use a method for assessing convergence of
 an N-dimensional quantity or transform our set of coordinates into a scalar
 metric. The latter approach has been developed considerably over the last decade
 and a method now commonly used in Monte Carlo eigenvalue calculations is to use
-a metric called the <a class="reference external" href="https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-06-3737_entropy.pdf">Shannon entropy</a>, a concept borrowed from information
+a metric called the <a class="reference external" href="https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-06-3737.pdf">Shannon entropy</a>, a concept borrowed from information
 theory.</p>
 <p>To compute the Shannon entropy of the source distribution, we first need to
 discretize the source distribution rather than having a set of coordinates in
@@ -210,7 +210,7 @@ Convergence,&#8221; <em>Trans. Am. Nucl. Soc.</em>, <strong>98</strong>, 512 (20
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/geometry.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -54,7 +54,7 @@
 <h2>2.1. Constructive Solid Geometry<a class="headerlink" href="#constructive-solid-geometry" title="Permalink to this headline">¶</a></h2>
 <p>OpenMC uses a technique known as <a class="reference external" href="http://en.wikipedia.org/wiki/Constructive_solid_geometry">constructive solid geometry</a> (CSG) to build
 arbitrarily complex three-dimensional models in Euclidean space. In a CSG model,
-every unique object is described as the union, intersection, or difference of
+every unique object is described as the union and/or intersection of
 <em>half-spaces</em> created by bounding <a class="reference external" href="http://en.wikipedia.org/wiki/Surface">surfaces</a>. Every surface divides all of
 space into exactly two half-spaces. We can mathematically define a surface as a
 collection of points that satisfy an equation of the form <img class="math" src="../_images/math/a196569f2f3e96bf3fd428db69ffaa94df4d6e75.png" alt="f(x,y,z) = 0"/>
@@ -76,21 +76,20 @@ with radius <img class="math" src="../_images/math/9d86170e7de539c0ff999de09621e
 <p>In OpenMC, every surface defined by the user is assigned an integer to uniquely
 identify it. We can then refer to either of the two half-spaces created by a
 surface by a combination of the unique ID of the surface and a positive/negative
-sign.  Figure <a href="#fig-halfspace">3</a> shows an example of an ellipse with unique ID 1
+sign.  Figure <a href="#fig-halfspace">14</a> shows an example of an ellipse with unique ID 1
 dividing space into two half-spaces.</p>
 <div class="align-center figure" id="id7">
-<span id="fig-halfspace"></span><img src="../_images/halfspace.svg" /><p class="caption"><span class="caption-text">Figure 3: Example of an ellipse and its associated half-spaces.</span></p>
+<span id="fig-halfspace"></span><img src="../_images/halfspace.svg" /><p class="caption"><span class="caption-text">Figure 14: Example of an ellipse and its associated half-spaces.</span></p>
 </div>
 <p>References to half-spaces created by surfaces are used to define regions of
-space of uniform composition, known as <em>cells</em>. While some codes allow regions
-to be defined by intersections, unions, and differences or half-spaces, OpenMC
-is currently limited to cells defined only as intersections of
-half-spaces. Thus, the specification of the cell must include a list of
-half-space references whose intersection defines the region. The region is then
-assigned a material defined elsewhere. Figure <a href="#fig-union">4</a> shows an
-example of a cell defined as the intersection of an ellipse and two planes.</p>
+space of uniform composition, which are then assigned to <em>cells</em>. OpenMC allows
+regions to be defined using union, intersection, and complement operators. As in
+<a class="reference external" href="http://mcnp.lanl.gov">MCNP</a>, the intersection operator is implicit as doesn&#8217;t need to be written in a
+region specification. A defined region is then associated with a material
+composition in a cell. Figure <a href="#fig-union">15</a> shows an example of a cell region
+defined as the intersection of an ellipse and two planes.</p>
 <div class="align-center figure" id="id8">
-<span id="fig-union"></span><img src="../_images/union.svg" /><p class="caption"><span class="caption-text">Figure 4: The shaded region represents a cell bounded by three surfaces.</span></p>
+<span id="fig-union"></span><img src="../_images/union.svg" /><p class="caption"><span class="caption-text">Figure 15: The shaded region represents a cell bounded by three surfaces.</span></p>
 </div>
 <p>The ability to form regions based on bounding quadratic surfaces enables OpenMC
 to model arbitrarily complex three-dimensional objects. In practice, one is
@@ -192,6 +191,16 @@ z_0 \; R^2"/></td>
 <td><img class="math" src="../_images/math/399b4f4cbcdbbe1bed6094340454ce80a0e2429d.png" alt="x_0 \; y_0 \;
 z_0 \; R^2"/></td>
 </tr>
+<tr class="row-odd"><td>General quadric
+surface</td>
+<td>quadric</td>
+<td><img class="math" src="../_images/math/2542fd9fe743f3d9862754ecf1a9647846ca5b61.png" alt="Ax^2 + By^2 + Cz^2 +
+Dxy + Eyz + Fxz + Gx + Hy +
+Jz + K"/></td>
+<td><img class="math" src="../_images/math/d44aec50d29458b96c09e92c97104e670710c0f8.png" alt="A \; B \; C \; D
+\; E \; F \; G \; H \;
+J \; K"/></td>
+</tr>
 </tbody>
 </table>
 <div class="section" id="universes">
@@ -403,12 +412,12 @@ the particle is referred to as &#8220;indexing&#8221;.</p>
 <h3>2.4.1. Rectilinear Lattice Indexing<a class="headerlink" href="#rectilinear-lattice-indexing" title="Permalink to this headline">¶</a></h3>
 <p>Indices are assigned to tiles in a rectilinear lattice based on the tile&#8217;s
 position along the <img class="math" src="../_images/math/188c175aac0a8a9c22499336711b5d7256407254.png" alt="x"/>, <img class="math" src="../_images/math/b124ff74afb0914bb434e8fb849eb56d734412f8.png" alt="y"/>, and <img class="math" src="../_images/math/84d7271dd9e78c1e05d6c3c6ecb60309ef7dfc73.png" alt="z"/> axes.  Figure
-<a href="#fig-rect-lat">5</a> maps the indices for a 2D lattice.  The indices, (1, 1),
+<a href="#fig-rect-lat">16</a> maps the indices for a 2D lattice.  The indices, (1, 1),
 map to the lower-left tile.  (5, 1) and (5, 5) map to the lower-right and
 upper-right tiles, respectively.</p>
 <div class="align-center figure" id="id10">
 <span id="fig-rect-lat"></span><a class="reference internal image-reference" href="../_images/rect_lat.svg"><img src="../_images/rect_lat.svg" width="400px" /></a>
-<p class="caption"><span class="caption-text">Figure 5: Rectilinear lattice tile indices.</span></p>
+<p class="caption"><span class="caption-text">Figure 16: Rectilinear lattice tile indices.</span></p>
 </div>
 <p>In general, a lattice tile is specified by the three indices,
 <img class="math" src="../_images/math/5594347865c09b52b42e6af2c5881703ed18887e.png" alt="(i_x, i_y, i_z)"/>.  If a particle&#8217;s current coordinates are
@@ -428,13 +437,13 @@ corner of the lattice, and <img class="math" src="../_images/math/0007256293ab5c
 <p>A skewed coordinate system is used for indexing hexagonal lattice tiles.
 Rather than a <img class="math" src="../_images/math/b124ff74afb0914bb434e8fb849eb56d734412f8.png" alt="y"/>-axis, another axis is used that is rotated 30 degrees
 counter-clockwise from the <img class="math" src="../_images/math/b124ff74afb0914bb434e8fb849eb56d734412f8.png" alt="y"/>-axis.  This axis is referred to as the
-<img class="math" src="../_images/math/ad59b6e24a4a00ac621801f8d7513d68be654ab5.png" alt="\alpha"/>-axis.  Figure <a href="#fig-hex-lat">6</a> shows how 2D hexagonal tiles
+<img class="math" src="../_images/math/ad59b6e24a4a00ac621801f8d7513d68be654ab5.png" alt="\alpha"/>-axis.  Figure <a href="#fig-hex-lat">17</a> shows how 2D hexagonal tiles
 are mapped with the <img class="math" src="../_images/math/ef86b8c9ffd4486aedde641af8c4579872ff85e8.png" alt="(x, \alpha)"/> basis.  In this system, (0, 0) maps to
 the center tile, (0, 2) to the top tile, and (2, -1) to the middle tile on the
 right side.</p>
 <div class="align-center figure" id="id11">
 <span id="fig-hex-lat"></span><a class="reference internal image-reference" href="../_images/hex_lat.svg"><img src="../_images/hex_lat.svg" width="400px" /></a>
-<p class="caption"><span class="caption-text">Figure 6: Hexagonal lattice tile indices.</span></p>
+<p class="caption"><span class="caption-text">Figure 17: Hexagonal lattice tile indices.</span></p>
 </div>
 <p>Unfortunately, the indices cannot be determined with one simple formula as
 before.  Indexing requires a two-step process, a coarse step which determines a
@@ -697,7 +706,7 @@ w' = w + \frac{2 (\bar{x}u + \bar{y}v - R^2\bar{z}w)}{R^2 (1 + R^2) \bar{z}}"/><
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/index.html

@@ -14,7 +14,7 @@
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -25,7 +25,7 @@
     <script type="text/javascript" src="../_static/doctools.js"></script>
     <link rel="top" title="OpenMC Documentation" href="../index.html" />
     <link rel="next" title="1. Introduction" href="introduction.html" />
-    <link rel="prev" title="Release Notes for OpenMC 0.7.0" href="../releasenotes.html" /> 
+    <link rel="prev" title="Release Notes for OpenMC 0.7.1" href="../releasenotes.html" /> 
   </head>
   <body role="document">
       <div class="header" role="banner">
@@ -36,7 +36,7 @@
       <div class="topnav" role="navigation" aria-label="top navigation">
       
         <p>
-        «&#160;&#160;<a href="../releasenotes.html">Release Notes for OpenMC 0.7.0</a>
+        «&#160;&#160;<a href="../releasenotes.html">Release Notes for OpenMC 0.7.1</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -115,7 +115,11 @@
 </ul>
 </li>
 <li class="toctree-l2"><a class="reference internal" href="physics.html#transforming-a-particle-s-coordinates">5.8. Transforming a Particle&#8217;s Coordinates</a></li>
-<li class="toctree-l2"><a class="reference internal" href="physics.html#effect-of-thermal-motion-on-cross-sections">5.9. Effect of Thermal Motion on Cross Sections</a></li>
+<li class="toctree-l2"><a class="reference internal" href="physics.html#effect-of-thermal-motion-on-cross-sections">5.9. Effect of Thermal Motion on Cross Sections</a><ul>
+<li class="toctree-l3"><a class="reference internal" href="physics.html#constant-cross-section-model">5.9.1. Constant Cross Section Model</a></li>
+<li class="toctree-l3"><a class="reference internal" href="physics.html#energy-dependent-cross-section-model">5.9.2. Energy-Dependent Cross Section Model</a></li>
+</ul>
+</li>
 <li class="toctree-l2"><a class="reference internal" href="physics.html#sab-tables">5.10. S(<img class="math" src="../_images/math/35e1676e6106b7f88449b57d632777294e05c3a6.png" alt="\alpha,\beta,T"/>) Tables</a><ul>
 <li class="toctree-l3"><a class="reference internal" href="physics.html#calculating-integrated-cross-sections">5.10.1. Calculating Integrated Cross Sections</a></li>
 <li class="toctree-l3"><a class="reference internal" href="physics.html#outgoing-angle-for-coherent-elastic-scattering">5.10.2. Outgoing Angle for Coherent Elastic Scattering</a></li>
@@ -186,7 +190,7 @@
       <div class="bottomnav" role="navigation" aria-label="bottom navigation">
       
         <p>
-        «&#160;&#160;<a href="../releasenotes.html">Release Notes for OpenMC 0.7.0</a>
+        «&#160;&#160;<a href="../releasenotes.html">Release Notes for OpenMC 0.7.1</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="../index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -198,7 +202,7 @@
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/introduction.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -177,7 +177,7 @@ and its variance is calculated.</li>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/parallelization.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -102,7 +102,7 @@ in the case of an eigenvalue calculation). This idea is illustrated in
 <a class="reference internal" href="#figure-master-slave"><span>Communication pattern in master-slave algorithm.</span></a>.</p>
 <div class="align-center figure" id="id4">
 <span id="figure-master-slave"></span><img alt="../_images/master-slave.png" src="../_images/master-slave.png" />
-<p class="caption"><span class="caption-text">Figure 7: Communication pattern in master-slave algorithm.</span></p>
+<p class="caption"><span class="caption-text">Figure 18: Communication pattern in master-slave algorithm.</span></p>
 </div>
 <p>Eigenvalue calculations are slightly more difficult to parallelize than fixed
 source calculations since it is necessary to converge on the fission source
@@ -148,7 +148,7 @@ to an algorithm whereby the compute nodes communicate with one another only as
 needed. This concept is illustrated in <a class="reference internal" href="#figure-nearest-neighbor"><span>Communication pattern in nearest neighbor algorithm.</span></a>.</p>
 <div class="align-center figure" id="id5">
 <span id="figure-nearest-neighbor"></span><img alt="../_images/nearest-neighbor.png" src="../_images/nearest-neighbor.png" />
-<p class="caption"><span class="caption-text">Figure 8: Communication pattern in nearest neighbor algorithm.</span></p>
+<p class="caption"><span class="caption-text">Figure 19: Communication pattern in nearest neighbor algorithm.</span></p>
 </div>
 <p>Since the source sites for each cycle are sampled from the fission sites banked
 from the previous cycle, it is a common occurrence for a fission site to be
@@ -218,7 +218,7 @@ hatching, and the cross-hatched regions represent source sites that are
 communicated between adjacent nodes.</p>
 <div class="align-center figure" id="id6">
 <span id="figure-neighbor-example"></span><img alt="../_images/nearest-neighbor-example.png" src="../_images/nearest-neighbor-example.png" />
-<p class="caption"><span class="caption-text">Figure 9: Example of nearest neighbor algorithm.</span></p>
+<p class="caption"><span class="caption-text">Figure 20: Example of nearest neighbor algorithm.</span></p>
 </div>
 </div>
 <div class="section" id="cost-of-master-slave-algorithm">
@@ -522,7 +522,7 @@ Radiation Penetration Calculations on a Parallel Computer,&#8221;
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/physics.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -174,11 +174,11 @@ secondary photons from nuclear de-excitation are tracked in OpenMC.</p>
 <div class="section" id="reactions">
 <h2>5.5. <img class="math" src="../_images/math/86812743fd2241a6d3211913c9f5c801e4abde62.png" alt="(n,xn)"/> Reactions<a class="headerlink" href="#reactions" title="Permalink to this headline">¶</a></h2>
 <p>These types of reactions are just treated as inelastic scattering and as such
-are subject to the same procedure as described in
-<a class="reference internal" href="#inelastic-scatter"><span>Inelastic Scattering</span></a>. Rather than tracking multiple secondary neutrons, the
-weight of the outgoing neutron is multiplied by the number of secondary
-neutrons, e.g. for <img class="math" src="../_images/math/b51c1d14d772823c103e00abf9e448770f391c44.png" alt="(n,2n)"/>, only one outgoing neutron is tracked but its
-weight is doubled.</p>
+are subject to the same procedure as described in <a class="reference internal" href="#inelastic-scatter"><span>Inelastic Scattering</span></a>. For
+reactions with integral multiplicity, e.g., <img class="math" src="../_images/math/b51c1d14d772823c103e00abf9e448770f391c44.png" alt="(n,2n)"/>, an appropriate
+number of secondary neutrons are created. For reactions that have a multiplicity
+given as a function of the incoming neutron energy (which occasionally occurs
+for MT=5), the weight of the outgoing neutron is multiplied by the multiplcity.</p>
 </div>
 <div class="section" id="fission">
 <span id="id1"></span><h2>5.6. Fission<a class="headerlink" href="#fission" title="Permalink to this headline">¶</a></h2>
@@ -763,14 +763,17 @@ probability distribution function can be found by integrating equation
 d\mathbf{v}_T \, R(\mathbf{v}_T)}"/></p>
 </div><p>Let us call the normalization factor in the denominator of equation
 <a href="#equation-target-pdf-1">(61)</a> <img class="math" src="../_images/math/2bcc65482aa8e15cd4c9e9f2542451fb4e971a91.png" alt="C"/>.</p>
-<p>It is normally assumed that <img class="math" src="../_images/math/15c26742484427f2a5cca0695fb67068ccc7437d.png" alt="\sigma (v_r)"/> is constant over the range of
+<div class="section" id="constant-cross-section-model">
+<h3>5.9.1. Constant Cross Section Model<a class="headerlink" href="#constant-cross-section-model" title="Permalink to this headline">¶</a></h3>
+<p>It is often assumed that <img class="math" src="../_images/math/15c26742484427f2a5cca0695fb67068ccc7437d.png" alt="\sigma (v_r)"/> is constant over the range of
 relative velocities of interest. This is a good assumption for almost all cases
 since the elastic scattering cross section varies slowly with velocity for light
 nuclei, and for heavy nuclei where large variations can occur due to resonance
 scattering, the moderating effect is rather small. Nonetheless, this assumption
 may cause incorrect answers in systems with low-lying resonances that can cause
 a significant amount of up-scatter that would be ignored by this assumption
-(e.g. U-238 in commercial light-water reactors). Nevertheless, with this
+(e.g. U-238 in commercial light-water reactors). We will revisit this assumption
+later in <a class="reference internal" href="#energy-dependent-xs-model"><span>Energy-Dependent Cross Section Model</span></a>. For now, continuing with the
 assumption, we write <img class="math" src="../_images/math/10031c0879c1d72e62b6e4be71acdb3bb845af55.png" alt="\sigma (v_r) = \sigma_s"/> which simplifies
 <a href="#equation-target-pdf-1">(61)</a> to</p>
 <div class="math" id="equation-target-pdf-2">
@@ -891,6 +894,26 @@ on the unit interval. Since the maximum value of <img class="math" src="../_imag
 cosine until a combination is found that satisfies equation
 <a href="#equation-freegas-accept-2">(80)</a>.</p>
 </div>
+<div class="section" id="energy-dependent-cross-section-model">
+<span id="energy-dependent-xs-model"></span><h3>5.9.2. Energy-Dependent Cross Section Model<a class="headerlink" href="#energy-dependent-cross-section-model" title="Permalink to this headline">¶</a></h3>
+<p>As was noted earlier, assuming that the elastic scattering cross section is
+constant in <a href="#equation-reaction-rate">(60)</a> is not strictly correct, especially when
+low-lying resonances are present in the cross sections for heavy nuclides. To
+correctly account for energy dependence of the scattering cross section entails
+performing another rejection step. The most common method is to sample
+<img class="math" src="../_images/math/126e84ba38f7dece5f0ad64e929b9588b20f6440.png" alt="\mu"/> and <img class="math" src="../_images/math/576d4929ebf1fbb6d919095c4e9d38a115c5d23a.png" alt="v_T"/> as in the constant cross section approximation and
+then perform a rejection on the ratio of the 0 K elastic scattering cross
+section at the relative velocity to the maximum 0 K elastic scattering cross
+section over the range of velocities considered:</p>
+<div class="math" id="equation-dbrc">
+<p><span class="eqno">(81)</span><img src="../_images/math/6069c1d194d08831dac6d14c6c98a4ff476dca4a.png" alt="p_{dbrc} = \frac{\sigma_s(v_r)}{\sigma_{s,max}}"/></p>
+</div><p>where it should be noted that the maximum is taken over the range <img class="math" src="../_images/math/5c4fc0ac8edd6ce5b9093890970e60ae4d760c73.png" alt="[v_n -
+4/\beta, 4_n + 4\beta]"/>. This method is known as Doppler broadening rejection
+correction (DBRC) and was first introduced by <a class="reference external" href="http://dx.doi.org/10.1016/j.anucene.2008.12.001">Becker et al.</a>. OpenMC has an
+implementation of DBRC as well as an accelerated sampling method that are
+described fully in <a class="reference external" href="http://dx.doi.org/10.1016/j.anucene.2014.01.017">Walsh et al.</a></p>
+</div>
+</div>
 <div class="section" id="sab-tables">
 <span id="id6"></span><h2>5.10. S(<img class="math" src="../_images/math/35e1676e6106b7f88449b57d632777294e05c3a6.png" alt="\alpha,\beta,T"/>) Tables<a class="headerlink" href="#sab-tables" title="Permalink to this headline">¶</a></h2>
 <p>For neutrons with thermal energies, generally less than 4 eV, the kinematics of
@@ -939,14 +962,14 @@ scattering, the cross sections are stored as linearly interpolable functions on
 a specified energy grid. For coherent elastic data, the cross section can be
 expressed as</p>
 <div class="math" id="equation-coherent-elastic-xs">
-<p><span class="eqno">(81)</span><img src="../_images/math/69b9fd7e970f8fe8b5426d25c6e19496d623eb9c.png" alt="\sigma(E) = \frac{\sigma_c}{E} \sum_{E_i &lt; E} f_i e^{-4WE_i}"/></p>
+<p><span class="eqno">(82)</span><img src="../_images/math/69b9fd7e970f8fe8b5426d25c6e19496d623eb9c.png" alt="\sigma(E) = \frac{\sigma_c}{E} \sum_{E_i &lt; E} f_i e^{-4WE_i}"/></p>
 </div><p>where <img class="math" src="../_images/math/58798ce83b38f451070ce0e5345d8913710b83fb.png" alt="\sigma_c"/> is the effective bound coherent scattering cross section,
 <img class="math" src="../_images/math/fa34489cf430b7630b30f28b2eb050e2489e3523.png" alt="W"/> is the effective Debye-Waller coefficient, <img class="math" src="../_images/math/b027139ad09fca3ca4adfe7d1149b9b2a69dbe3a.png" alt="E_i"/> are the
 energies of the Bragg edges, and <img class="math" src="../_images/math/f3e8f82a52bdd47c816d880bdbe685ecfc88160b.png" alt="f_i"/> are related to crystallographic
 structure factors. Since the functional form of the cross section is just 1/E
 and the proportionality constant changes only at Bragg edges, the
 proportionality constants are stored and then the cross section can be
-calculated analytically based on equation <a href="#equation-coherent-elastic-xs">(81)</a>.</p>
+calculated analytically based on equation <a href="#equation-coherent-elastic-xs">(82)</a>.</p>
 </div>
 <div class="section" id="outgoing-angle-for-coherent-elastic-scattering">
 <h3>5.10.2. Outgoing Angle for Coherent Elastic Scattering<a class="headerlink" href="#outgoing-angle-for-coherent-elastic-scattering" title="Permalink to this headline">¶</a></h3>
@@ -957,11 +980,11 @@ does change. For coherent elastic scattering, the angle will depend on which
 Bragg edge scattered the neutron. The probability that edge <img class="math" src="../_images/math/a581f053bbfa5115f42c13094857cdd12a37ec49.png" alt="i"/> will
 scatter then neutron is given by</p>
 <div class="math" id="equation-coherent-elastic-probability">
-<p><span class="eqno">(82)</span><img src="../_images/math/7f5e985e66252bad6ff62ab015cbbc5a15e60018.png" alt="\frac{f_i e^{-4WE_i}}{\sum_j f_j e^{-4WE_j}}."/></p>
+<p><span class="eqno">(83)</span><img src="../_images/math/7f5e985e66252bad6ff62ab015cbbc5a15e60018.png" alt="\frac{f_i e^{-4WE_i}}{\sum_j f_j e^{-4WE_j}}."/></p>
 </div><p>After a Bragg edge has been sampled, the cosine of the angle of scattering is
 given analytically by</p>
 <div class="math" id="equation-coherent-elastic-angle">
-<p><span class="eqno">(83)</span><img src="../_images/math/3946360f436553e959b22b8ad46e35c7367f8fa9.png" alt="\mu = 1 - \frac{E_i}{E}"/></p>
+<p><span class="eqno">(84)</span><img src="../_images/math/3946360f436553e959b22b8ad46e35c7367f8fa9.png" alt="\mu = 1 - \frac{E_i}{E}"/></p>
 </div><p>where <img class="math" src="../_images/math/b027139ad09fca3ca4adfe7d1149b9b2a69dbe3a.png" alt="E_i"/> is the energy of the Bragg edge that scattered the neutron.</p>
 </div>
 <div class="section" id="outgoing-angle-for-incoherent-elastic-scattering">
@@ -973,10 +996,10 @@ elastic energy grid. First the outgoing angle bin <img class="math" src="../_ima
 the incoming energy of the neutron satisfies <img class="math" src="../_images/math/11bed1d62b2f553c6bba9255ade361778cb870ea.png" alt="E_i &lt; E &lt; E_{i+1}"/> the final
 cosine is</p>
 <div class="math" id="equation-incoherent-elastic-angle">
-<p><span class="eqno">(84)</span><img src="../_images/math/69363585db6d8c715083681d2511f1ffcefadf04.png" alt="\mu = \mu_{i,j} + f (\mu_{i+1,j} - \mu_{i,j})"/></p>
+<p><span class="eqno">(85)</span><img src="../_images/math/69363585db6d8c715083681d2511f1ffcefadf04.png" alt="\mu = \mu_{i,j} + f (\mu_{i+1,j} - \mu_{i,j})"/></p>
 </div><p>where the interpolation factor is defined as</p>
 <div class="math" id="equation-sab-interpolation-factor">
-<p><span class="eqno">(85)</span><img src="../_images/math/3fbc578039b24cebbb2f87a16a6cc65b0d3f50a2.png" alt="f = \frac{E - E_i}{E_{i+1} - E_i}."/></p>
+<p><span class="eqno">(86)</span><img src="../_images/math/3fbc578039b24cebbb2f87a16a6cc65b0d3f50a2.png" alt="f = \frac{E - E_i}{E_{i+1} - E_i}."/></p>
 </div></div>
 <div class="section" id="outgoing-energy-and-angle-for-inelastic-scattering">
 <h3>5.10.4. Outgoing Energy and Angle for Inelastic Scattering<a class="headerlink" href="#outgoing-energy-and-angle-for-inelastic-scattering" title="Permalink to this headline">¶</a></h3>
@@ -994,17 +1017,17 @@ discrete equiprobable outgoing cosines.</p>
 outgoing energy spectra is represented in the ACE data as a set of discrete and
 equiprobable outgoing energies.  The procedure to determine the outgoing energy
 and angle is as such. First, the interpolation factor is determined from
-equation <a href="#equation-sab-interpolation-factor">(85)</a>.  Then, an outgoing energy bin is
+equation <a href="#equation-sab-interpolation-factor">(86)</a>.  Then, an outgoing energy bin is
 sampled from a uniform distribution and then interpolated between values
 corresponding to neighboring incoming energies:</p>
 <div class="math" id="equation-inelastic-energy">
-<p><span class="eqno">(86)</span><img src="../_images/math/7321eed5feceac9bc2a006d3372a7504afec4e7f.png" alt="E = E_{i,j} + f (E_{i+1,j} - E_{i,j})"/></p>
+<p><span class="eqno">(87)</span><img src="../_images/math/7321eed5feceac9bc2a006d3372a7504afec4e7f.png" alt="E = E_{i,j} + f (E_{i+1,j} - E_{i,j})"/></p>
 </div><p>where <img class="math" src="../_images/math/2222b995c2353461d6e0ac399df3bf276b9c7304.png" alt="E_{i,j}"/> is the j-th outgoing energy corresponding to the i-th
 incoming energy. For each combination of incoming and outgoing energies, there
 is a series equiprobable outgoing cosines. An outgoing cosine bin is sampled
 uniformly and then the final cosine is interpolated on the incoming energy grid:</p>
 <div class="math" id="equation-inelastic-angle">
-<p><span class="eqno">(87)</span><img src="../_images/math/81c9ecfb768181e525934dce274b01155d16656a.png" alt="\mu = \mu_{i,j,k} + f (\mu_{i+1,j,k} - \mu_{i,j,k})"/></p>
+<p><span class="eqno">(88)</span><img src="../_images/math/81c9ecfb768181e525934dce274b01155d16656a.png" alt="\mu = \mu_{i,j,k} + f (\mu_{i+1,j,k} - \mu_{i,j,k})"/></p>
 </div><p>where <img class="math" src="../_images/math/42f768e135c2ac3dc468b2985fbdfc750dc5f0c2.png" alt="\mu_{i,j,k}"/> is the k-th outgoing cosine corresponding to the j-th
 outgoing energy and the i-th incoming energy.</p>
 </div>
@@ -1035,7 +1058,7 @@ which angular distribution data to use.  Like the linear-linear interpolation
 case in Law 61, the angular distribution closest to the sampled value of the
 cumulative distribution function for the outgoing energy is utilized.  The
 actual algorithm utilized to sample the outgoing angle is shown in equation
-<a href="#equation-inelastic-angle">(87)</a>.</p>
+<a href="#equation-inelastic-angle">(88)</a>.</p>
 </div>
 </div>
 </div>
@@ -1077,17 +1100,17 @@ capture cross sections from the probability tables interpolating between
 neighboring incoming energies. If interpolation is specified, then
 the cross sections are calculated as</p>
 <div class="math" id="equation-ptables-linlin">
-<p><span class="eqno">(88)</span><img src="../_images/math/ea9369891f1ff8938ae390bdb648678e78c3003a.png" alt="\sigma = \sigma_{i,j} + f (\sigma_{i+1,j} - \sigma{i,j})"/></p>
+<p><span class="eqno">(89)</span><img src="../_images/math/ea9369891f1ff8938ae390bdb648678e78c3003a.png" alt="\sigma = \sigma_{i,j} + f (\sigma_{i+1,j} - \sigma{i,j})"/></p>
 </div><p>where <img class="math" src="../_images/math/fbbd7cbe39753a1f3c7fd7d850bb95bd97781a6c.png" alt="\sigma_{i,j}"/> is the j-th band cross section corresponding to the
 i-th incoming neutron energy and <img class="math" src="../_images/math/0001d02b63ede2fe3219e05a7cd09c82ae6298b6.png" alt="f"/> is the interpolation factor defined
-in the same manner as <a href="#equation-sab-interpolation-factor">(85)</a>. If logarithmic
+in the same manner as <a href="#equation-sab-interpolation-factor">(86)</a>. If logarithmic
 interpolation is specified, the cross sections are calculated as</p>
 <div class="math" id="equation-ptables-loglog">
-<p><span class="eqno">(89)</span><img src="../_images/math/0af49ad5ceac407e47acc6bc425307523ccb028f.png" alt="\sigma = \exp \left ( \log \sigma_{i,j} + f \log
+<p><span class="eqno">(90)</span><img src="../_images/math/0af49ad5ceac407e47acc6bc425307523ccb028f.png" alt="\sigma = \exp \left ( \log \sigma_{i,j} + f \log
 \frac{\sigma_{i+1,j}}{\sigma_{i,j}} \right )"/></p>
 </div><p>where the interpolation factor is now defined as</p>
 <div class="math" id="equation-log-interpolation-factor">
-<p><span class="eqno">(90)</span><img src="../_images/math/acb86d5cd33acbbd379014727295964c15cef03a.png" alt="f = \frac{\log \frac{E}{E_i}}{\log \frac{E_{i+1}}{E_i}}."/></p>
+<p><span class="eqno">(91)</span><img src="../_images/math/acb86d5cd33acbbd379014727295964c15cef03a.png" alt="f = \frac{\log \frac{E}{E_i}}{\log \frac{E_{i+1}}{E_i}}."/></p>
 </div><p>A flag is also present in the probability table that specifies whether an
 inelastic cross section should be calculated. If so, this is done from a normal
 reaction cross section (either MT=51 or a special MT). Finally, if the
@@ -1110,7 +1133,7 @@ this is a misnomer) is commonly used.</p>
 instead, at every collision, the weight of neutron is reduced by probability of
 absorption occurring, i.e.</p>
 <div class="math" id="equation-survival-biasing-weight">
-<p><span class="eqno">(91)</span><img src="../_images/math/7309379b4001d8097c2b5bc82a4145d140ca0404.png" alt="w' = w \left ( 1 - \frac{\sigma_a (E)}{\sigma_t (E)} \right )"/></p>
+<p><span class="eqno">(92)</span><img src="../_images/math/7309379b4001d8097c2b5bc82a4145d140ca0404.png" alt="w' = w \left ( 1 - \frac{\sigma_a (E)}{\sigma_t (E)} \right )"/></p>
 </div><p>where <img class="math" src="../_images/math/d0f641ddabfc77111d0823470829ca6d9ca59ad1.png" alt="w'"/> is the weight of the neutron after adjustment and <img class="math" src="../_images/math/8659700e6646cd91bc02c32affaa5ec046ee9935.png" alt="w"/> is
 the weight of the neutron before adjustment. A few other things need to be
 handled differently if survival biasing is turned on. Although fission reactions
@@ -1120,7 +1143,7 @@ successive generations. The algorithm for sampling fission sites is the same as
 that described in <a class="reference internal" href="#fission"><span>Fission</span></a>. The only difference is in equation
 <a href="#equation-fission-neutrons">(14)</a>. We now need to produce</p>
 <div class="math" id="equation-fission-neutrons-survival">
-<p><span class="eqno">(92)</span><img src="../_images/math/23e01bdec04e2cbbc81a53da4451026f6a4c6287.png" alt="\nu = \frac{w}{k} \frac{\nu_t \sigma_f(E)}{\sigma_t (E)}"/></p>
+<p><span class="eqno">(93)</span><img src="../_images/math/23e01bdec04e2cbbc81a53da4451026f6a4c6287.png" alt="\nu = \frac{w}{k} \frac{\nu_t \sigma_f(E)}{\sigma_t (E)}"/></p>
 </div><p>fission sites, where <img class="math" src="../_images/math/8659700e6646cd91bc02c32affaa5ec046ee9935.png" alt="w"/> is the weight of the neutron before being
 adjusted. One should note this is just the expected number of neutrons produced
 <em>per collision</em> rather than the expected number of neutrons produced given that
@@ -1211,7 +1234,7 @@ book can be obtained for free from the <a class="reference external" href="http:
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/random_numbers.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
         COLLAPSE_INDEX: false,
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         HAS_SOURCE:  true
@@ -124,7 +124,7 @@ Different Sizes and Good Lattice Structures,&#8221; <em>Math. Comput.</em>, <str
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

methods/tallies.html

@@ -14,7 +14,7 @@
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       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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@@ -421,7 +421,7 @@ in Statistics - Simulation and Computation, 16 (4), pp. 1123-1132 (1987).</td></
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
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 <script type="text/javascript">
 

publications.html

@@ -14,7 +14,7 @@
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -25,7 +25,7 @@
     <script type="text/javascript" src="_static/doctools.js"></script>
     <link rel="top" title="OpenMC Documentation" href="index.html" />
     <link rel="next" title="License Agreement" href="license.html" />
-    <link rel="prev" title="Tally Arithmetic" href="pythonapi/examples/tally-arithmetic.html" /> 
+    <link rel="prev" title="MGXS Part III: Libraries" href="pythonapi/examples/mgxs-part-iii.html" /> 
   </head>
   <body role="document">
       <div class="header" role="banner">
@@ -36,7 +36,7 @@
       <div class="topnav" role="navigation" aria-label="top navigation">
       
         <p>
-        «&#160;&#160;<a href="pythonapi/examples/tally-arithmetic.html">Tally Arithmetic</a>
+        «&#160;&#160;<a href="pythonapi/examples/mgxs-part-iii.html">MGXS Part III: Libraries</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -68,6 +68,9 @@ Code,&#8221; <em>Ann. Nucl. Energy</em>, <strong>51</strong>, 274&#8211;281
 <div class="section" id="benchmarking">
 <h2>Benchmarking<a class="headerlink" href="#benchmarking" title="Permalink to this headline">¶</a></h2>
 <ul class="simple">
+<li>Khurrum S. Chaudri and Sikander M. Mirza, &#8220;Burnup dependent Monte Carlo
+neutron physics calculations of IAEA MTR benchmark,&#8221; <em>Prog. Nucl. Energy</em>,
+<strong>81</strong>, 43-52 (2015). <a class="reference external" href="http://dx.doi.org/j.pnucene.2014.12.018">http://dx.doi.org/j.pnucene.2014.12.018</a></li>
 <li>Daniel J. Kelly, Brian N. Aviles, Paul K. Romano, Bryan R. Herman,
 Nicholas E. Horelik, and Benoit Forget, &#8220;Analysis of select BEAVRS PWR
 benchmark cycle 1 results using MC21 and OpenMC,&#8221; <em>Proc. PHYSOR</em>, Kyoto,
@@ -95,12 +98,8 @@ simulations,&#8221; <em>Proc. Joint Int. Conf. M&amp;C+SNA+MC</em>, Nashville, T
 Apr. 19&#8211;23 (2015).</li>
 <li>Bryan R. Herman, Benoit Forget, and Kord Smith, &#8220;Progress toward Monte
 Carlo-thermal hydraulic coupling using low-order nonlinear diffusion
-acceleration methods.&#8221; In press, <em>Ann. Nucl. Energy</em>,
-(2014). <a class="reference external" href="http://dx.doi.org/10.1016/j.anucene/2014.10.029">http://dx.doi.org/10.1016/j.anucene/2014.10.029</a></li>
-<li>Adam G. Nelson and William R. Martin, &#8220;Improved Convergence of Monte Carlo
-Generated Multi-Group Scattering Moments,&#8221; <em>Proc. Int. Conf. Mathematics and
-Computational Methods Applied to Nuclear Science and Engineering</em>, Sun Valley,
-Idaho, May 5&#8211;9 (2013).</li>
+acceleration methods.&#8221; <em>Ann. Nucl. Energy</em>, <strong>84</strong>, 63-72
+(2015). <a class="reference external" href="http://dx.doi.org/10.1016/j.anucene.2014.10.029">http://dx.doi.org/10.1016/j.anucene.2014.10.029</a></li>
 <li>Bryan R. Herman, Benoit Forget, and Kord Smith, &#8220;Utilizing CMFD in OpenMC to
 Estimate Dominance Ratio and Adjoint,&#8221; <em>Trans. Am. Nucl. Soc.</em>, <strong>109</strong>,
 1389-1392 (2013).</li>
@@ -117,18 +116,55 @@ solid geometries,&#8221; <em>Proc. PHYSOR</em>, Kyoto, Japan, Sep. 28&#8211;Oct.
 <div class="section" id="miscellaneous">
 <h2>Miscellaneous<a class="headerlink" href="#miscellaneous" title="Permalink to this headline">¶</a></h2>
 <ul class="simple">
+<li>William Boyd, Sterling Harper, and Paul K. Romano, &#8220;Equipping OpenMC for the
+big data era,&#8221; Accepted, <em>PHYSOR 2016</em>, Sun Valley, Idaho, May 1-5, 2016.</li>
+<li>Qicang Shen, William Boyd, Benoit Forget, and Kord Smith, &#8220;Tally precision
+triggers for the OpenMC Monte Carlo code,&#8221; <em>Trans. Am. Nucl. Soc.</em>, <strong>112</strong>,
+637-640 (2015).</li>
 <li>Timothy P. Burke, Brian C. Kiedrowski, and William R. Martin, &#8220;Flux and
 Reaction Rate Kernel Density Estimators in OpenMC,&#8221; <em>Trans. Am. Nucl. Soc.</em>,
 <strong>109</strong>, 683-686 (2013).</li>
 </ul>
 </div>
+<div class="section" id="multi-group-cross-section-generation">
+<h2>Multi-group Cross Section Generation<a class="headerlink" href="#multi-group-cross-section-generation" title="Permalink to this headline">¶</a></h2>
+<ul class="simple">
+<li>Adam G. Nelson and William R. Martin, &#8220;Improved Monte Carlo tallying of
+multi-group scattering moments using the NDPP code,&#8221; <em>Trans. Am. Nucl. Soc.</em>,
+<strong>113</strong>, 645-648 (2015)</li>
+<li>Adam G. Nelson and William R. Martin, &#8220;Improved Monte Carlo tallying of
+multi-group scattering moment matrices,&#8221; <em>Trans. Am. Nucl. Soc.</em>, <strong>110</strong>,
+217-220 (2014).</li>
+<li>Adam G. Nelson and William R. Martin, &#8220;Improved Convergence of Monte Carlo
+Generated Multi-Group Scattering Moments,&#8221; <em>Proc. Int. Conf. Mathematics and
+Computational Methods Applied to Nuclear Science and Engineering</em>, Sun Valley,
+Idaho, May 5&#8211;9 (2013).</li>
+</ul>
+</div>
 <div class="section" id="nuclear-data">
 <h2>Nuclear Data<a class="headerlink" href="#nuclear-data" title="Permalink to this headline">¶</a></h2>
 <ul class="simple">
+<li>Colin Josey, Pablo Ducru, Benoit Forget, and Kord Smith, &#8220;Windowed multipole
+for cross section Doppler broadening,&#8221; <em>J. Comput. Phys.</em>, In Press
+(2016). <a class="reference external" href="http://dx.doi.org/10.1016/jcp.2015.08.013">http://dx.doi.org/10.1016/jcp.2015.08.013</a></li>
+<li>Colin Josey, Benoit Forget, and Kord Smith, &#8220;Windowed multipole sensitivity to
+target accuracy of the optimization procedure,&#8221; <em>J. Nucl. Sci. Technol.</em>,
+<strong>52</strong>, 987-992 (2015). <a class="reference external" href="http://dx.doi.org/10.1080/00223131.2015.1035353">http://dx.doi.org/10.1080/00223131.2015.1035353</a></li>
+<li>Jonathan A. Walsh, Paul K. Romano, Benoit Forget, and Kord S. Smith,
+&#8220;Optimizations of the energy grid search algorithm in continuous-energy Monte
+Carlo particle transport codes&#8221;, <em>Comput. Phys. Commun.</em>, <strong>196</strong>, 134-142
+(2015). <a class="reference external" href="http://dx.doi.org/10.1016/j.cpc.2015.05.025">http://dx.doi.org/10.1016/j.cpc.2015.05.025</a></li>
 <li>Jonathan A. Walsh, Benoit Forget, Kord S. Smith, Brian C. Kiedrowski, and
 Forrest B. Brown, &#8220;Direct, on-the-fly calculation of unresolved resonance
 region cross sections in Monte Carlo simulations,&#8221; <em>Proc. Joint
 Int. Conf. M&amp;C+SNA+MC</em>, Nashville, Tennessee, Apr. 19&#8211;23 (2015).</li>
+<li>Amanda L. Lund, Andrew R. Siegel, Benoit Forget, Colin Josey, and
+Paul K. Romano, &#8220;Using fractional cascading to accelerate cross section
+lookups in Monte Carlo particle transport calculations,&#8221; <em>Proc. Joint
+Int. Conf. M&amp;C+SNA+MC</em>, Nashville, Tennessee, Apr. 19&#8211;23 (2015).</li>
+<li>Ronald O. Rahaman, Andrew R. Siegel, and Paul K. Romano, &#8220;Monte Carlo
+performance analysis for varying cross section parameter regimes,&#8221;
+<em>Proc. Joint Int. Conf. M&amp;C+SNA+MC</em>, Nashville, Tennessee, Apr. 19&#8211;23 (2015).</li>
 <li>Paul K. Romano and Timothy H. Trumbull, &#8220;Comparison of algorithms for Doppler
 broadening pointwise tabulated cross sections,&#8221; <em>Ann. Nucl. Energy</em>, <strong>75</strong>,
 358&#8211;364 (2015). <a class="reference external" href="http://dx.doi.org/10.1016/j.anucene.2014.08.046">http://dx.doi.org/10.1016/j.anucene.2014.08.046</a></li>
@@ -146,6 +182,9 @@ Carlo simulations using the multipole representation,&#8221; <em>Ann. Nucl. Ener
 <div class="section" id="parallelism">
 <h2>Parallelism<a class="headerlink" href="#parallelism" title="Permalink to this headline">¶</a></h2>
 <ul class="simple">
+<li>Paul K. Romano, John R. Tramm, and Andrew R. Siegel, &#8220;Efficacy of hardware
+threading for Monte Carlo particle transport calculations on multi- and
+many-core systems,&#8221; Accepted, <em>PHYSOR 2016</em>, Sun Valley, Idaho, May 1-5, 2016.</li>
 <li>David Ozog, Allen D. Malony, and Andrew R. Siegel, &#8220;A performance analysis of
 SIMD algorithms for Monte Carlo simulations of nuclear reactor cores,&#8221;
 <em>Proc. IEEE Int. Parallel and Distributed Processing Symposium</em>, Hyderabad,
@@ -207,7 +246,7 @@ Carlo Criticality Calculations,&#8221; <em>Nucl. Sci. Eng.</em>, <strong>170</st
       <div class="bottomnav" role="navigation" aria-label="bottom navigation">
       
         <p>
-        «&#160;&#160;<a href="pythonapi/examples/tally-arithmetic.html">Tally Arithmetic</a>
+        «&#160;&#160;<a href="pythonapi/examples/mgxs-part-iii.html">MGXS Part III: Libraries</a>
         &#160;&#160;::&#160;&#160;
         <a class="uplink" href="index.html">Contents</a>
         &#160;&#160;::&#160;&#160;
@@ -219,7 +258,7 @@ Carlo Criticality Calculations,&#8221; <em>Nucl. Sci. Eng.</em>, <strong>170</st
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

py-modindex.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    './',
-        VERSION:     '0.7.0',
+        VERSION:     '0.7.1',
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         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -101,6 +101,16 @@
        <td>&nbsp;&nbsp;&nbsp;
        <a href="pythonapi/mesh.html#module-openmc.mesh"><code class="xref">openmc.mesh</code></a></td><td>
        <em></em></td></tr>
+     <tr class="cg-1">
+       <td></td>
+       <td>&nbsp;&nbsp;&nbsp;
+       <a href="pythonapi/energy_groups.html#module-openmc.mgxs.groups"><code class="xref">openmc.mgxs.groups</code></a></td><td>
+       <em></em></td></tr>
+     <tr class="cg-1">
+       <td></td>
+       <td>&nbsp;&nbsp;&nbsp;
+       <a href="pythonapi/mgxs_library.html#module-openmc.mgxs.library"><code class="xref">openmc.mgxs.library</code></a></td><td>
+       <em></em></td></tr>
      <tr class="cg-1">
        <td></td>
        <td>&nbsp;&nbsp;&nbsp;
@@ -171,7 +181,7 @@
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

pythonapi/ace.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
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+        VERSION:     '0.7.1',
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         FILE_SUFFIX: '.html',
         HAS_SOURCE:  true
@@ -87,7 +87,7 @@
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

pythonapi/cmfd.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
         URL_ROOT:    '../',
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -308,7 +308,7 @@ must be used to identify any fission source region.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

pythonapi/element.html

@@ -14,7 +14,7 @@
     <script type="text/javascript">
       var DOCUMENTATION_OPTIONS = {
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+        VERSION:     '0.7.1',
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         HAS_SOURCE:  true
@@ -82,6 +82,13 @@ abundances.</p>
 <p>Cross section identifier, e.g. 71c</p>
 </dd></dl>
 
+<dl class="attribute">
+<dt id="openmc.element.Element.scattering">
+<code class="descname">scattering</code><a class="headerlink" href="#openmc.element.Element.scattering" title="Permalink to this definition">¶</a></dt>
+<dd><p><em>&#8216;data&#8217; or &#8216;iso-in-lab&#8217; or None</em></p>
+<p>The type of angular scattering distribution to use</p>
+</dd></dl>
+
 </dd></dl>
 
 </div>
@@ -103,7 +110,7 @@ abundances.</p>
 
     <div class="footer" role="contentinfo">
         &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
-      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.1.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
     </div>
 <script type="text/javascript">
 

pythonapi/energy_groups.html

@@ -0,0 +1,211 @@
+<!DOCTYPE html PUBLIC "-//W3C//DTD XHTML 1.0 Transitional//EN"
+  "http://www.w3.org/TR/xhtml1/DTD/xhtml1-transitional.dtd">
+
+
+<html xmlns="http://www.w3.org/1999/xhtml">
+  <head>
+    <meta http-equiv="Content-Type" content="text/html; charset=utf-8" />
+    
+    <title>Energy Groups &mdash; OpenMC Documentation</title>
+    
+    <link rel="stylesheet" href="../_static/haiku.css" type="text/css" />
+    <link rel="stylesheet" href="../_static/pygments.css" type="text/css" />
+    
+    <script type="text/javascript">
+      var DOCUMENTATION_OPTIONS = {
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+        HAS_SOURCE:  true
+      };
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+    <link rel="top" title="OpenMC Documentation" href="../index.html" />
+    <link rel="up" title="Python API" href="index.html" />
+    <link rel="next" title="MGXS Library" href="mgxs_library.html" />
+    <link rel="prev" title="Multi-Group Cross Sections" href="mgxs.html" /> 
+  </head>
+  <body role="document">
+      <div class="header" role="banner">
+        <a href="../index.html">
+          <img class="logo" src="../_static/openmc.png" alt="Logo"/>
+        </a>
+      </div>
+      <div class="topnav" role="navigation" aria-label="top navigation">
+      
+        <p>
+        «&#160;&#160;<a href="mgxs.html">Multi-Group Cross Sections</a>
+        &#160;&#160;::&#160;&#160;
+        <a class="uplink" href="../index.html">Contents</a>
+        &#160;&#160;::&#160;&#160;
+        <a href="mgxs_library.html">MGXS Library</a>&#160;&#160;»
+        </p>
+
+      </div>
+      <div class="content">
+        
+        
+  <div class="section" id="module-openmc.mgxs.groups">
+<span id="energy-groups"></span><span id="pythonapi-energy-groups"></span><h1>Energy Groups<a class="headerlink" href="#module-openmc.mgxs.groups" title="Permalink to this headline">¶</a></h1>
+<dl class="class">
+<dt id="openmc.mgxs.groups.EnergyGroups">
+<em class="property">class </em><code class="descclassname">openmc.mgxs.groups.</code><code class="descname">EnergyGroups</code><span class="sig-paren">(</span><em>group_edges=None</em><span class="sig-paren">)</span><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups" title="Permalink to this definition">¶</a></dt>
+<dd><p>An energy groups structure used for multi-group cross-sections.</p>
+<table class="docutils field-list" frame="void" rules="none">
+<col class="field-name" />
+<col class="field-body" />
+<tbody valign="top">
+<tr class="field-odd field"><th class="field-name">Parameters:</th><td class="field-body"><strong>group_edges</strong> (<em>Iterable of Real</em>) &#8211; The energy group boundaries [MeV]</td>
+</tr>
+</tbody>
+</table>
+<dl class="attribute">
+<dt id="openmc.mgxs.groups.EnergyGroups.group_edges">
+<code class="descname">group_edges</code><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.group_edges" title="Permalink to this definition">¶</a></dt>
+<dd><p><em>Iterable of Real</em></p>
+<p>The energy group boundaries [MeV]</p>
+</dd></dl>
+
+<dl class="attribute">
+<dt id="openmc.mgxs.groups.EnergyGroups.num_group">
+<code class="descname">num_group</code><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.num_group" title="Permalink to this definition">¶</a></dt>
+<dd><p><em>Integral</em></p>
+<p>The number of energy groups</p>
+</dd></dl>
+
+<dl class="method">
+<dt id="openmc.mgxs.groups.EnergyGroups.get_condensed_groups">
+<code class="descname">get_condensed_groups</code><span class="sig-paren">(</span><em>coarse_groups</em><span class="sig-paren">)</span><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.get_condensed_groups" title="Permalink to this definition">¶</a></dt>
+<dd><p>Return a coarsened version of this EnergyGroups object.</p>
+<p>This method merges together energy groups in this object into wider
+energy groups as defined by the list of groups specified by the user,
+and returns a new, coarse EnergyGroups object.</p>
+<table class="docutils field-list" frame="void" rules="none">
+<col class="field-name" />
+<col class="field-body" />
+<tbody valign="top">
+<tr class="field-odd field"><th class="field-name">Parameters:</th><td class="field-body"><strong>coarse_groups</strong> (<em>Iterable of 2-tuple</em>) &#8211; The energy groups of interest - a list of 2-tuples, each directly
+corresponding to one of the new coarse groups. The values in the
+2-tuples are upper/lower energy groups used to construct a new
+coarse group. For example, if [(1,2), (3,4)] was used as the coarse
+groups, fine groups 1 and 2 would be merged into coarse group 1
+while fine groups 3 and 4 would be merged into coarse group 2.</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Returns:</th><td class="field-body">A coarsened version of this EnergyGroups object.</td>
+</tr>
+<tr class="field-odd field"><th class="field-name">Return type:</th><td class="field-body"><a class="reference internal" href="#openmc.mgxs.groups.EnergyGroups" title="openmc.mgxs.groups.EnergyGroups">EnergyGroups</a></td>
+</tr>
+<tr class="field-even field"><th class="field-name">Raises:</th><td class="field-body"><code class="xref py py-exc docutils literal"><span class="pre">ValueError</span></code> &#8211;
+If the group edges have not yet been set.</td>
+</tr>
+</tbody>
+</table>
+</dd></dl>
+
+<dl class="method">
+<dt id="openmc.mgxs.groups.EnergyGroups.get_group">
+<code class="descname">get_group</code><span class="sig-paren">(</span><em>energy</em><span class="sig-paren">)</span><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.get_group" title="Permalink to this definition">¶</a></dt>
+<dd><p>Returns the energy group in which the given energy resides.</p>
+<table class="docutils field-list" frame="void" rules="none">
+<col class="field-name" />
+<col class="field-body" />
+<tbody valign="top">
+<tr class="field-odd field"><th class="field-name">Parameters:</th><td class="field-body"><strong>energy</strong> (<em>Real</em>) &#8211; The energy of interest in MeV</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Returns:</th><td class="field-body">The energy group index, starting at 1 for the highest energies</td>
+</tr>
+<tr class="field-odd field"><th class="field-name">Return type:</th><td class="field-body">Integral</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Raises:</th><td class="field-body"><code class="xref py py-exc docutils literal"><span class="pre">ValueError</span></code> &#8211;
+If the group edges have not yet been set.</td>
+</tr>
+</tbody>
+</table>
+</dd></dl>
+
+<dl class="method">
+<dt id="openmc.mgxs.groups.EnergyGroups.get_group_bounds">
+<code class="descname">get_group_bounds</code><span class="sig-paren">(</span><em>group</em><span class="sig-paren">)</span><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.get_group_bounds" title="Permalink to this definition">¶</a></dt>
+<dd><p>Returns the energy boundaries for the energy group of interest.</p>
+<table class="docutils field-list" frame="void" rules="none">
+<col class="field-name" />
+<col class="field-body" />
+<tbody valign="top">
+<tr class="field-odd field"><th class="field-name">Parameters:</th><td class="field-body"><strong>group</strong> (<em>Integral</em>) &#8211; The energy group index, starting at 1 for the highest energies</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Returns:</th><td class="field-body">The low and high energy bounds for the group in MeV</td>
+</tr>
+<tr class="field-odd field"><th class="field-name">Return type:</th><td class="field-body">2-tuple</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Raises:</th><td class="field-body"><code class="xref py py-exc docutils literal"><span class="pre">ValueError</span></code> &#8211;
+If the group edges have not yet been set.</td>
+</tr>
+</tbody>
+</table>
+</dd></dl>
+
+<dl class="method">
+<dt id="openmc.mgxs.groups.EnergyGroups.get_group_indices">
+<code class="descname">get_group_indices</code><span class="sig-paren">(</span><em>groups='all'</em><span class="sig-paren">)</span><a class="headerlink" href="#openmc.mgxs.groups.EnergyGroups.get_group_indices" title="Permalink to this definition">¶</a></dt>
+<dd><p>Returns the array indices for one or more energy groups.</p>
+<table class="docutils field-list" frame="void" rules="none">
+<col class="field-name" />
+<col class="field-body" />
+<tbody valign="top">
+<tr class="field-odd field"><th class="field-name">Parameters:</th><td class="field-body"><strong>groups</strong> (<em>str, tuple</em>) &#8211; The energy groups of interest - a tuple of the energy group indices,
+starting at 1 for the highest energies (default is &#8216;all&#8217;)</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Returns:</th><td class="field-body">The ndarray array indices for each energy group of interest</td>
+</tr>
+<tr class="field-odd field"><th class="field-name">Return type:</th><td class="field-body">ndarray</td>
+</tr>
+<tr class="field-even field"><th class="field-name">Raises:</th><td class="field-body"><code class="xref py py-exc docutils literal"><span class="pre">ValueError</span></code> &#8211;
+If the group edges have not yet been set, or if a group is requested
+that is outside the bounds of the number of energy groups.</td>
+</tr>
+</tbody>
+</table>
+</dd></dl>
+
+</dd></dl>
+
+</div>
+
+
+      </div>
+      <div class="bottomnav" role="navigation" aria-label="bottom navigation">
+      
+        <p>
+        «&#160;&#160;<a href="mgxs.html">Multi-Group Cross Sections</a>
+        &#160;&#160;::&#160;&#160;
+        <a class="uplink" href="../index.html">Contents</a>
+        &#160;&#160;::&#160;&#160;
+        <a href="mgxs_library.html">MGXS Library</a>&#160;&#160;»
+        </p>
+
+      </div>
+
+
+    <div class="footer" role="contentinfo">
+        &copy; Copyright 2011-2015, Massachusetts Institute of Technology.
+      Created using <a href="http://sphinx-doc.org/">Sphinx</a> 1.3.3.
+    </div>
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+</html>

pythonapi/examples/mgxs-part-i-content.html

@@ -0,0 +1,846 @@
+<!DOCTYPE html PUBLIC "-//W3C//DTD XHTML 1.0 Transitional//EN"
+  "http://www.w3.org/TR/xhtml1/DTD/xhtml1-transitional.dtd">
+
+
+<html xmlns="http://www.w3.org/1999/xhtml">
+  <head>
+    <meta http-equiv="Content-Type" content="text/html; charset=utf-8" />
+    
+    <title>Introduction to Multi-Group Cross Sections (MGXS) &mdash; OpenMC Documentation</title>
+    
+    <link rel="stylesheet" href="../../_static/haiku.css" type="text/css" />
+    <link rel="stylesheet" href="../../_static/pygments.css" type="text/css" />
+    
+    <script type="text/javascript">
+      var DOCUMENTATION_OPTIONS = {
+        URL_ROOT:    '../../',
+        VERSION:     '0.7.0',
+        COLLAPSE_INDEX: false,
+        FILE_SUFFIX: '.html',
+        HAS_SOURCE:  true
+      };
+    </script>
+    <script type="text/javascript" src="../../_static/jquery.js"></script>
+    <script type="text/javascript" src="../../_static/underscore.js"></script>
+    <script type="text/javascript" src="../../_static/doctools.js"></script>
+    <script type="text/javascript" src="https://cdn.mathjax.org/mathjax/latest/MathJax.js?config=TeX-AMS-MML_HTMLorMML"></script>
+    <link rel="top" title="OpenMC Documentation" href="../../index.html" /> 
+  </head>
+  <body role="document">
+      <div class="header" role="banner">
+        <a href="../../index.html">
+          <img class="logo" src="../../_static/openmc.png" alt="Logo"/>
+        </a>
+      </div>
+      <div class="topnav" role="navigation" aria-label="top navigation">
+      
+        <p>
+        <a class="uplink" href="../../index.html">Contents</a>
+        </p>
+
+      </div>
+      <div class="content">
+        
+        
+  <p>This IPython Notebook introduces the use of the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> module
+to calculate multi-group cross sections for an infinite homogeneous
+medium. In particular, this Notebook introduces the the following
+features:</p>
+<ul class="simple">
+<li><strong>General equations</strong> for scalar-flux averaged multi-group cross
+sections</li>
+<li>Creation of multi-group cross sections for an <strong>infinite homogeneous
+medium</strong></li>
+<li>Use of <strong>tally arithmetic</strong> to manipulate multi-group cross sections</li>
+</ul>
+<p><strong>Note:</strong> This Notebook illustrates the use of
+<a class="reference external" href="http://pandas.pydata.org/">Pandas</a> <code class="docutils literal"><span class="pre">DataFrames</span></code> to containerize
+multi-group cross section data. We recommend using
+<a class="reference external" href="http://pandas.pydata.org/">Pandas</a> &gt;v0.15.0 or later since OpenMC&#8217;s
+Python API leverages the multi-indexing feature included in the most
+recent releases of <a class="reference external" href="http://pandas.pydata.org/">Pandas</a>.</p>
+<div class="section" id="introduction-to-multi-group-cross-sections-mgxs">
+<h1>Introduction to Multi-Group Cross Sections (MGXS)<a class="headerlink" href="#introduction-to-multi-group-cross-sections-mgxs" title="Permalink to this headline">¶</a></h1>
+<p>Many Monte Carlo particle transport codes, including OpenMC, use
+continuous-energy nuclear cross section data. However, most
+deterministic neutron transport codes use <em>multi-group cross sections</em>
+defined over discretized energy bins or <em>energy groups</em>. An example of
+U-235&#8217;s continuous-energy fission cross section along with a 16-group
+cross section computed for a light water reactor spectrum is displayed
+below.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="kn">from</span> <span class="nn">IPython.display</span> <span class="kn">import</span> <span class="n">Image</span>
+<span class="n">Image</span><span class="p">(</span><span class="n">filename</span><span class="o">=</span><span class="s">&#39;images/mgxs.png&#39;</span><span class="p">,</span> <span class="n">width</span><span class="o">=</span><span class="mi">350</span><span class="p">)</span>
+</pre></div>
+</div>
+<img alt="../../_images/mgxs-part-i-content_3_0.png" src="../../_images/mgxs-part-i-content_3_0.png" />
+<p>A variety of tools employing different methodologies have been developed
+over the years to compute multi-group cross sections for certain
+applications, including NJOY (LANL), MC<span class="math">\(^2\)</span>-3 (ANL), and Serpent
+(VTT). The <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> Python module is designed to leverage
+OpenMC&#8217;s tally system to calculate multi-group cross sections with
+arbitrary energy discretizations for fine-mesh heterogeneous
+deterministic neutron transport applications.</p>
+<p>Before proceeding to illustrate how one may use the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code>
+module, it is worthwhile to define the general equations used to
+calculate multi-group cross sections. This is only intended as a brief
+overview of the methodology used by <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> - we refer the
+interested reader to the large body of literature on the subject for a
+more comprehensive understanding of this complex topic.</p>
+<div class="section" id="introductory-notation">
+<h2>Introductory Notation<a class="headerlink" href="#introductory-notation" title="Permalink to this headline">¶</a></h2>
+<p>The continuous real-valued microscopic cross section may be denoted
+<span class="math">\(\sigma_{n,x}(\mathbf{r}, E)\)</span> for position vector
+<span class="math">\(\mathbf{r}\)</span>, energy <span class="math">\(E\)</span>, nuclide <span class="math">\(n\)</span> and interaction
+type <span class="math">\(x\)</span>. Similarly, the scalar neutron flux may be denoted by
+<span class="math">\(\Phi(\mathbf{r},E)\)</span> for position <span class="math">\(\mathbf{r}\)</span> and energy
+<span class="math">\(E\)</span>. <strong>Note</strong>: Although nuclear cross sections are dependent on
+the temperature <span class="math">\(T\)</span> of the interacting medium, the temperature
+variable is neglected here for brevity.</p>
+</div>
+<div class="section" id="spatial-and-energy-discretization">
+<h2>Spatial and Energy Discretization<a class="headerlink" href="#spatial-and-energy-discretization" title="Permalink to this headline">¶</a></h2>
+<p>The energy domain for critical systems such as thermal reactors spans
+more than 10 orders of magnitude of neutron energies from
+10<span class="math">\(^{-5}\)</span> - 10<span class="math">\(^7\)</span> eV. The multi-group approximation
+discretization divides this energy range into one or more energy groups.
+In particular, for <span class="math">\(G\)</span> total groups, we denote an energy group
+index <span class="math">\(g\)</span> such that <span class="math">\(g \in \{1, 2, ..., G\}\)</span>. 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 <strong>energy
+condensation</strong>.</p>
+<p>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 <span class="math">\(K\)</span> distinct spatial zones, we designate each
+spatial zone an index <span class="math">\(k\)</span> such that
+<span class="math">\(k \in \{1, 2, ..., K\}\)</span>. The volume of each spatial zone is
+denoted by <span class="math">\(V_{k}\)</span>. The integration over discrete spatial zones is
+commonly referred to as <strong>spatial homogenization</strong>.</p>
+</div>
+<div class="section" id="general-scalar-flux-weighted-mgxs">
+<h2>General Scalar-Flux Weighted MGXS<a class="headerlink" href="#general-scalar-flux-weighted-mgxs" title="Permalink to this headline">¶</a></h2>
+<p>The multi-group cross sections computed by <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> are defined
+as a <em>scalar flux-weighted average</em> of the microscopic cross sections
+across each discrete energy group. This formulation is employed in order
+to preserve the reaction rates within each energy group and spatial
+zone. In particular, spatial homogenization and energy condensation are
+used to compute the general multi-group cross section
+<span class="math">\(\sigma_{n,x,k,g}\)</span> as follows:</p>
+<div class="math">
+\[\sigma_{n,x,k,g} = \frac{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\sigma_{n,x}(\mathbf{r},E')\Phi(\mathbf{r},E')}{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\Phi(\mathbf{r},E')}\]</div>
+<p>This scalar flux-weighted average microscopic cross section is computed
+by <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> for most multi-group cross sections, including total,
+absorption, and fission reaction types. These double integrals are
+stochastically computed with OpenMC&#8217;s tally system - in particular,
+<a class="reference external" href="https://mit-crpg.github.io/openmc/pythonapi/filter.html">filters</a> on
+the energy range and spatial zone (material, cell or universe) define
+the bounds of integration for both numerator and denominator.</p>
+</div>
+<div class="section" id="multi-group-scattering-matrices">
+<h2>Multi-Group Scattering Matrices<a class="headerlink" href="#multi-group-scattering-matrices" title="Permalink to this headline">¶</a></h2>
+<p>The general multi-group cross section <span class="math">\(\sigma_{n,x,k,g}\)</span> is a
+vector of <span class="math">\(G\)</span> values for each energy group <span class="math">\(g\)</span>. The equation
+presented above only discretizes the energy of the incoming neutron and
+neglects the outgoing energy of the neutron (if any). Hence, this
+formulation must be extended to account for the outgoing energy of
+neutrons in the discretized scattering matrix cross section used by
+deterministic neutron transport codes.</p>
+<p>We denote the incoming and outgoing neutron energy groups as <span class="math">\(g\)</span>
+and <span class="math">\(g'\)</span> for the microscopic scattering matrix cross section
+<span class="math">\(\sigma_{n,s}(\mathbf{r},E)\)</span>. As before, spatial homogenization
+and energy condensation are used to find the multi-group scattering
+matrix cross section <span class="math">\(\sigma_{n,s,k,g \to g'}\)</span> as follows:</p>
+<div class="math">
+\[\sigma_{n,s,k,g\rightarrow g'} = \frac{\int_{E_{g'}}^{E_{g'-1}}\mathrm{d}E''\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\sigma_{n,s}(\mathbf{r},E'\rightarrow E'')\Phi(\mathbf{r},E')}{\int_{E_{g}}^{E_{g-1}}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\Phi(\mathbf{r},E')}\]</div>
+<p>This scalar flux-weighted multi-group microscopic scattering matrix is
+computed using OpenMC tallies with both energy in and energy out
+filters.</p>
+</div>
+<div class="section" id="multi-group-fission-spectrum">
+<h2>Multi-Group Fission Spectrum<a class="headerlink" href="#multi-group-fission-spectrum" title="Permalink to this headline">¶</a></h2>
+<p>The energy spectrum of neutrons emitted from fission is denoted by
+<span class="math">\(\chi_{n}(\mathbf{r},E' \rightarrow E'')\)</span> for incoming and
+outgoing energies <span class="math">\(E'\)</span> and <span class="math">\(E''\)</span>, respectively. Unlike the
+multi-group cross sections <span class="math">\(\sigma_{n,x,k,g}\)</span> considered up to
+this point, the fission spectrum is a probability distribution and must
+sum to unity. The outgoing energy is typically much less dependent on
+the incoming energy for fission than for scattering interactions. As a
+result, it is common practice to integrate over the incoming neutron
+energy when computing the multi-group fission spectrum. The fission
+spectrum may be simplified as <span class="math">\(\chi_{n}(\mathbf{r},E)\)</span> with
+outgoing energy <span class="math">\(E\)</span>.</p>
+<p>Unlike the multi-group cross sections defined up to this point, the
+multi-group fission spectrum is weighted by the fission production rate
+rather than the scalar flux. This formulation is intended to preserve
+the total fission production rate in the multi-group deterministic
+calculation. In order to mathematically define the multi-group fission
+spectrum, we denote the microscopic fission cross section as
+<span class="math">\(\sigma_{n,f}(\mathbf{r},E)\)</span> and the average number of neutrons
+emitted from fission interactions with nuclide <span class="math">\(n\)</span> as
+<span class="math">\(\nu_{n}(\mathbf{r},E)\)</span>. The multi-group fission spectrum
+<span class="math">\(\chi_{n,k,g}\)</span> is then the probability of fission neutrons emitted
+into energy group <span class="math">\(g\)</span>.</p>
+<p>Similar to before, spatial homogenization and energy condensation are
+used to find the multi-group fission spectrum <span class="math">\(\chi_{n,k,g}\)</span> as
+follows:</p>
+<div class="math">
+\[\chi_{n,k,g'} = \frac{\int_{E_{g'}}^{E_{g'-1}}\mathrm{d}E''\int_{0}^{\infty}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\chi_{n}(\mathbf{r},E'\rightarrow E'')\nu_{n}(\mathbf{r},E')\sigma_{n,f}(\mathbf{r},E')\Phi(\mathbf{r},E')}{\int_{0}^{\infty}\mathrm{d}E'\int_{\mathbf{r} \in V_{k}}\mathrm{d}\mathbf{r}\nu_{n}(\mathbf{r},E')\sigma_{n,f}(\mathbf{r},E')\Phi(\mathbf{r},E')}\]</div>
+<p>The fission production-weighted multi-group fission spectrum is computed
+using OpenMC tallies with both energy in and energy out filters.</p>
+<p>This concludes our brief overview on the methodology to compute
+multi-group cross sections. The following sections detail more
+concretely how users may employ the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> module to power
+simulation workflows requiring multi-group cross sections for downstream
+deterministic calculations.</p>
+</div>
+</div>
+<div class="section" id="generate-input-files">
+<h1>Generate Input Files<a class="headerlink" href="#generate-input-files" title="Permalink to this headline">¶</a></h1>
+<div class="code python highlight-python"><div class="highlight"><pre>import numpy as np
+import matplotlib.pyplot as plt
+
+import openmc
+import openmc.mgxs as mgxs
+
+%matplotlib inline
+</pre></div>
+</div>
+<p>First we need to define materials that will be used in the problem.
+Before defining a material, we must create nuclides that are used in the
+material.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate some Nuclides</span>
+<span class="n">h1</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Nuclide</span><span class="p">(</span><span class="s">&#39;H-1&#39;</span><span class="p">)</span>
+<span class="n">o16</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Nuclide</span><span class="p">(</span><span class="s">&#39;O-16&#39;</span><span class="p">)</span>
+<span class="n">u235</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Nuclide</span><span class="p">(</span><span class="s">&#39;U-235&#39;</span><span class="p">)</span>
+<span class="n">u238</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Nuclide</span><span class="p">(</span><span class="s">&#39;U-238&#39;</span><span class="p">)</span>
+<span class="n">zr90</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Nuclide</span><span class="p">(</span><span class="s">&#39;Zr-90&#39;</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>With the nuclides we defined, we will now create a material for the
+homogeneous medium.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate a Material and register the Nuclides</span>
+<span class="n">inf_medium</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Material</span><span class="p">(</span><span class="n">name</span><span class="o">=</span><span class="s">&#39;moderator&#39;</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">set_density</span><span class="p">(</span><span class="s">&#39;g/cc&#39;</span><span class="p">,</span> <span class="mf">5.</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">add_nuclide</span><span class="p">(</span><span class="n">h1</span><span class="p">,</span>  <span class="mf">0.028999667</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">add_nuclide</span><span class="p">(</span><span class="n">o16</span><span class="p">,</span> <span class="mf">0.01450188</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">add_nuclide</span><span class="p">(</span><span class="n">u235</span><span class="p">,</span> <span class="mf">0.000114142</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">add_nuclide</span><span class="p">(</span><span class="n">u238</span><span class="p">,</span> <span class="mf">0.006886019</span><span class="p">)</span>
+<span class="n">inf_medium</span><span class="o">.</span><span class="n">add_nuclide</span><span class="p">(</span><span class="n">zr90</span><span class="p">,</span> <span class="mf">0.002116053</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>With our material, we can now create a <code class="docutils literal"><span class="pre">MaterialsFile</span></code> object that can
+be exported to an actual XML file.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate a MaterialsFile, register all Materials, and export to XML</span>
+<span class="n">materials_file</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">MaterialsFile</span><span class="p">()</span>
+<span class="n">materials_file</span><span class="o">.</span><span class="n">default_xs</span> <span class="o">=</span> <span class="s">&#39;71c&#39;</span>
+<span class="n">materials_file</span><span class="o">.</span><span class="n">add_material</span><span class="p">(</span><span class="n">inf_medium</span><span class="p">)</span>
+<span class="n">materials_file</span><span class="o">.</span><span class="n">export_to_xml</span><span class="p">()</span>
+</pre></div>
+</div>
+<p>Now let&#8217;s move on to the geometry. This problem will be a simple square
+cell with reflective boundary conditions to simulate an infinite
+homogeneous medium. The first step is to create the outer bounding
+surfaces of the problem.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate boundary Planes</span>
+<span class="n">min_x</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">XPlane</span><span class="p">(</span><span class="n">boundary_type</span><span class="o">=</span><span class="s">&#39;reflective&#39;</span><span class="p">,</span> <span class="n">x0</span><span class="o">=-</span><span class="mf">0.63</span><span class="p">)</span>
+<span class="n">max_x</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">XPlane</span><span class="p">(</span><span class="n">boundary_type</span><span class="o">=</span><span class="s">&#39;reflective&#39;</span><span class="p">,</span> <span class="n">x0</span><span class="o">=</span><span class="mf">0.63</span><span class="p">)</span>
+<span class="n">min_y</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">YPlane</span><span class="p">(</span><span class="n">boundary_type</span><span class="o">=</span><span class="s">&#39;reflective&#39;</span><span class="p">,</span> <span class="n">y0</span><span class="o">=-</span><span class="mf">0.63</span><span class="p">)</span>
+<span class="n">max_y</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">YPlane</span><span class="p">(</span><span class="n">boundary_type</span><span class="o">=</span><span class="s">&#39;reflective&#39;</span><span class="p">,</span> <span class="n">y0</span><span class="o">=</span><span class="mf">0.63</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>With the surfaces defined, we can now create a cell that is defined by
+intersections of half-spaces created by the surfaces.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate a Cell</span>
+<span class="n">cell</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Cell</span><span class="p">(</span><span class="n">cell_id</span><span class="o">=</span><span class="mi">1</span><span class="p">,</span> <span class="n">name</span><span class="o">=</span><span class="s">&#39;cell&#39;</span><span class="p">)</span>
+
+<span class="c"># Register bounding Surfaces with the Cell</span>
+<span class="n">cell</span><span class="o">.</span><span class="n">region</span> <span class="o">=</span> <span class="o">+</span><span class="n">min_x</span> <span class="o">&amp;</span> <span class="o">-</span><span class="n">max_x</span> <span class="o">&amp;</span> <span class="o">+</span><span class="n">min_y</span> <span class="o">&amp;</span> <span class="o">-</span><span class="n">max_y</span>
+
+<span class="c"># Fill the Cell with the Material</span>
+<span class="n">cell</span><span class="o">.</span><span class="n">fill</span> <span class="o">=</span> <span class="n">inf_medium</span>
+</pre></div>
+</div>
+<p>OpenMC requires that there is a &#8220;root&#8221; universe. Let us create a root
+universe and add our square cell to it.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate Universe</span>
+<span class="n">root_universe</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Universe</span><span class="p">(</span><span class="n">universe_id</span><span class="o">=</span><span class="mi">0</span><span class="p">,</span> <span class="n">name</span><span class="o">=</span><span class="s">&#39;root universe&#39;</span><span class="p">)</span>
+<span class="n">root_universe</span><span class="o">.</span><span class="n">add_cell</span><span class="p">(</span><span class="n">cell</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>We now must create a geometry that is assigned a root universe, put the
+geometry into a <code class="docutils literal"><span class="pre">GeometryFile</span></code> object, and export it to XML.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Create Geometry and set root Universe</span>
+<span class="n">openmc_geometry</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Geometry</span><span class="p">()</span>
+<span class="n">openmc_geometry</span><span class="o">.</span><span class="n">root_universe</span> <span class="o">=</span> <span class="n">root_universe</span>
+
+<span class="c"># Instantiate a GeometryFile</span>
+<span class="n">geometry_file</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">GeometryFile</span><span class="p">()</span>
+<span class="n">geometry_file</span><span class="o">.</span><span class="n">geometry</span> <span class="o">=</span> <span class="n">openmc_geometry</span>
+
+<span class="c"># Export to &quot;geometry.xml&quot;</span>
+<span class="n">geometry_file</span><span class="o">.</span><span class="n">export_to_xml</span><span class="p">()</span>
+</pre></div>
+</div>
+<p>Next, we must define simulation parameters. In this case, we will use 10
+inactive batches and 40 active batches each with 2500 particles.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># OpenMC simulation parameters</span>
+<span class="n">batches</span> <span class="o">=</span> <span class="mi">50</span>
+<span class="n">inactive</span> <span class="o">=</span> <span class="mi">10</span>
+<span class="n">particles</span> <span class="o">=</span> <span class="mi">2500</span>
+
+<span class="c"># Instantiate a SettingsFile</span>
+<span class="n">settings_file</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">SettingsFile</span><span class="p">()</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">batches</span> <span class="o">=</span> <span class="n">batches</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">inactive</span> <span class="o">=</span> <span class="n">inactive</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">particles</span> <span class="o">=</span> <span class="n">particles</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">output</span> <span class="o">=</span> <span class="p">{</span><span class="s">&#39;tallies&#39;</span><span class="p">:</span> <span class="bp">True</span><span class="p">,</span> <span class="s">&#39;summary&#39;</span><span class="p">:</span> <span class="bp">True</span><span class="p">}</span>
+<span class="n">bounds</span> <span class="o">=</span> <span class="p">[</span><span class="o">-</span><span class="mf">0.63</span><span class="p">,</span> <span class="o">-</span><span class="mf">0.63</span><span class="p">,</span> <span class="o">-</span><span class="mf">0.63</span><span class="p">,</span> <span class="mf">0.63</span><span class="p">,</span> <span class="mf">0.63</span><span class="p">,</span> <span class="mf">0.63</span><span class="p">]</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">set_source_space</span><span class="p">(</span><span class="s">&#39;fission&#39;</span><span class="p">,</span> <span class="n">bounds</span><span class="p">)</span>
+
+<span class="c"># Export to &quot;settings.xml&quot;</span>
+<span class="n">settings_file</span><span class="o">.</span><span class="n">export_to_xml</span><span class="p">()</span>
+</pre></div>
+</div>
+<p>Now we are ready to generate multi-group cross sections! First, let&#8217;s
+define a 2-group structure using the built-in <code class="docutils literal"><span class="pre">EnergyGroups</span></code> class.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate a 2-group EnergyGroups object</span>
+<span class="n">groups</span> <span class="o">=</span> <span class="n">mgxs</span><span class="o">.</span><span class="n">EnergyGroups</span><span class="p">()</span>
+<span class="n">groups</span><span class="o">.</span><span class="n">group_edges</span> <span class="o">=</span> <span class="n">np</span><span class="o">.</span><span class="n">array</span><span class="p">([</span><span class="mf">0.</span><span class="p">,</span> <span class="mf">0.625e-6</span><span class="p">,</span> <span class="mf">20.</span><span class="p">])</span>
+</pre></div>
+</div>
+<p>We can now use the <code class="docutils literal"><span class="pre">EnergyGroups</span></code> object, along with our previously
+created materials and geometry, to instantiate some <code class="docutils literal"><span class="pre">MGXS</span></code> objects
+from the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> module. In particular, the following are
+subclasses of the generic and abstract <code class="docutils literal"><span class="pre">MGXS</span></code> class:</p>
+<ul class="simple">
+<li><code class="docutils literal"><span class="pre">TotalXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">TransportXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">AbsorptionXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">CaptureXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">FissionXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">NuFissionXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">ScatterXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">NuScatterXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">ScatterMatrixXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">NuScatterMatrixXS</span></code></li>
+<li><code class="docutils literal"><span class="pre">Chi</span></code></li>
+</ul>
+<p>These classes provide us with an interface to generate the tally inputs
+as well as perform post-processing of OpenMC&#8217;s tally data to compute the
+respective multi-group cross sections. In this case, let&#8217;s create the
+multi-group total, absorption and scattering cross sections with our
+2-group structure.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate a few different sections</span>
+<span class="n">total</span> <span class="o">=</span> <span class="n">mgxs</span><span class="o">.</span><span class="n">TotalXS</span><span class="p">(</span><span class="n">domain</span><span class="o">=</span><span class="n">cell</span><span class="p">,</span> <span class="n">domain_type</span><span class="o">=</span><span class="s">&#39;cell&#39;</span><span class="p">,</span> <span class="n">groups</span><span class="o">=</span><span class="n">groups</span><span class="p">)</span>
+<span class="n">absorption</span> <span class="o">=</span> <span class="n">mgxs</span><span class="o">.</span><span class="n">AbsorptionXS</span><span class="p">(</span><span class="n">domain</span><span class="o">=</span><span class="n">cell</span><span class="p">,</span> <span class="n">domain_type</span><span class="o">=</span><span class="s">&#39;cell&#39;</span><span class="p">,</span> <span class="n">groups</span><span class="o">=</span><span class="n">groups</span><span class="p">)</span>
+<span class="n">scattering</span> <span class="o">=</span> <span class="n">mgxs</span><span class="o">.</span><span class="n">ScatterXS</span><span class="p">(</span><span class="n">domain</span><span class="o">=</span><span class="n">cell</span><span class="p">,</span> <span class="n">domain_type</span><span class="o">=</span><span class="s">&#39;cell&#39;</span><span class="p">,</span> <span class="n">groups</span><span class="o">=</span><span class="n">groups</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>Each multi-group cross section object stores its tallies in a Python
+dictionary called <code class="docutils literal"><span class="pre">tallies</span></code>. We can inspect the tallies in the
+dictionary for our <code class="docutils literal"><span class="pre">Absorption</span></code> object as follows.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="n">absorption</span><span class="o">.</span><span class="n">tallies</span>
+</pre></div>
+</div>
+<div class="highlight-python"><div class="highlight"><pre>OrderedDict([(&#39;flux&#39;, Tally
+    ID             =        10000
+    Name           =
+    Filters        =
+                            cell    [1]
+                            energy  [  0.00000000e+00   6.25000000e-07   2.00000000e+01]
+    Nuclides       =        total
+    Scores         =        [&#39;flux&#39;]
+    Estimator      =        tracklength
+), (&#39;absorption&#39;, Tally
+    ID             =        10001
+    Name           =
+    Filters        =
+                            cell    [1]
+                            energy  [  0.00000000e+00   6.25000000e-07   2.00000000e+01]
+    Nuclides       =        total
+    Scores         =        [&#39;absorption&#39;]
+    Estimator      =        tracklength
+)])
+</pre></div>
+</div>
+<p>The <code class="docutils literal"><span class="pre">Absorption</span></code> object includes tracklength tallies for the
+&#8216;absorption&#8217; and &#8216;flux&#8217; scores in the 2-group structure in cell 1. Now
+that each <code class="docutils literal"><span class="pre">MGXS</span></code> object contains the tallies that it needs, we must
+add these tallies to a <code class="docutils literal"><span class="pre">TalliesFile</span></code> object to generate the
+&#8220;tallies.xml&#8221; input file for OpenMC.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Instantiate an empty TalliesFile</span>
+<span class="n">tallies_file</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">TalliesFile</span><span class="p">()</span>
+
+<span class="c"># Add total tallies to the tallies file</span>
+<span class="k">for</span> <span class="n">tally</span> <span class="ow">in</span> <span class="n">total</span><span class="o">.</span><span class="n">tallies</span><span class="o">.</span><span class="n">values</span><span class="p">():</span>
+    <span class="n">tallies_file</span><span class="o">.</span><span class="n">add_tally</span><span class="p">(</span><span class="n">tally</span><span class="p">)</span>
+
+<span class="c"># Add absorption tallies to the tallies file</span>
+<span class="k">for</span> <span class="n">tally</span> <span class="ow">in</span> <span class="n">absorption</span><span class="o">.</span><span class="n">tallies</span><span class="o">.</span><span class="n">values</span><span class="p">():</span>
+    <span class="n">tallies_file</span><span class="o">.</span><span class="n">add_tally</span><span class="p">(</span><span class="n">tally</span><span class="p">)</span>
+
+<span class="c"># Add scattering tallies to the tallies file</span>
+<span class="k">for</span> <span class="n">tally</span> <span class="ow">in</span> <span class="n">scattering</span><span class="o">.</span><span class="n">tallies</span><span class="o">.</span><span class="n">values</span><span class="p">():</span>
+    <span class="n">tallies_file</span><span class="o">.</span><span class="n">add_tally</span><span class="p">(</span><span class="n">tally</span><span class="p">)</span>
+
+<span class="c"># Export to &quot;tallies.xml&quot;</span>
+<span class="n">tallies_file</span><span class="o">.</span><span class="n">export_to_xml</span><span class="p">()</span>
+</pre></div>
+</div>
+<p>Now we a have a complete set of inputs, so we can go ahead and run our
+simulation.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Run OpenMC</span>
+<span class="n">executor</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Executor</span><span class="p">()</span>
+<span class="n">executor</span><span class="o">.</span><span class="n">run_simulation</span><span class="p">()</span>
+</pre></div>
+</div>
+<div class="highlight-python"><div class="highlight"><pre>       .d88888b.                             888b     d888  .d8888b.
+      d88P&quot; &quot;Y88b                            8888b   d8888 d88P  Y88b
+      888     888                            88888b.d88888 888    888
+      888     888 88888b.   .d88b.  88888b.  888Y88888P888 888
+      888     888 888 &quot;88b d8P  Y8b 888 &quot;88b 888 Y888P 888 888
+      888     888 888  888 88888888 888  888 888  Y8P  888 888    888
+      Y88b. .d88P 888 d88P Y8b.     888  888 888   &quot;   888 Y88b  d88P
+       &quot;Y88888P&quot;  88888P&quot;   &quot;Y8888  888  888 888       888  &quot;Y8888P&quot;
+__________________888______________________________________________________
+                  888
+                  888
+
+      Copyright:      2011-2015 Massachusetts Institute of Technology
+      License:        http://mit-crpg.github.io/openmc/license.html
+      Version:        0.7.0
+      Git SHA1:       c4b14a5ef87f004528d35cbf33fef3ed15a386ca
+      Date/Time:      2015-12-02 09:11:05
+      MPI Processes:  1
+
+ ===========================================================================
+ ========================&gt;     INITIALIZATION     &lt;=========================
+ ===========================================================================
+
+ Reading settings XML file...
+ Reading cross sections XML file...
+ Reading geometry XML file...
+ Reading materials XML file...
+ Reading tallies XML file...
+ Building neighboring cells lists for each surface...
+ Loading ACE cross section table: 1001.71c
+ Loading ACE cross section table: 8016.71c
+ Loading ACE cross section table: 92235.71c
+ Loading ACE cross section table: 92238.71c
+ Loading ACE cross section table: 40090.71c
+ Maximum neutron transport energy: 20.0000 MeV for 1001.71c
+ Initializing source particles...
+
+ ===========================================================================
+ ====================&gt;     K EIGENVALUE SIMULATION     &lt;====================
+ ===========================================================================
+
+  Bat./Gen.      k            Average k
+  =========   ========   ====================
+        1/1    1.19804
+        2/1    1.12945
+        3/1    1.15573
+        4/1    1.13929
+        5/1    1.16300
+        6/1    1.22117
+        7/1    1.19012
+        8/1    1.11299
+        9/1    1.16066
+       10/1    1.12566
+       11/1    1.20854
+       12/1    1.14691    1.17773 +/- 0.03082
+       13/1    1.17204    1.17583 +/- 0.01789
+       14/1    1.14148    1.16724 +/- 0.01529
+       15/1    1.17272    1.16834 +/- 0.01189
+       16/1    1.18575    1.17124 +/- 0.01014
+       17/1    1.20498    1.17606 +/- 0.00983
+       18/1    1.14754    1.17249 +/- 0.00923
+       19/1    1.18141    1.17348 +/- 0.00820
+       20/1    1.15074    1.17121 +/- 0.00768
+       21/1    1.15914    1.17011 +/- 0.00703
+       22/1    1.14586    1.16809 +/- 0.00673
+       23/1    1.18999    1.16978 +/- 0.00642
+       24/1    1.15101    1.16844 +/- 0.00609
+       25/1    1.13791    1.16640 +/- 0.00602
+       26/1    1.19791    1.16837 +/- 0.00597
+       27/1    1.19818    1.17012 +/- 0.00587
+       28/1    1.14160    1.16854 +/- 0.00576
+       29/1    1.11487    1.16571 +/- 0.00614
+       30/1    1.17538    1.16620 +/- 0.00584
+       31/1    1.20210    1.16791 +/- 0.00581
+       32/1    1.20078    1.16940 +/- 0.00574
+       33/1    1.14624    1.16839 +/- 0.00558
+       34/1    1.14618    1.16747 +/- 0.00542
+       35/1    1.16866    1.16752 +/- 0.00520
+       36/1    1.18565    1.16821 +/- 0.00504
+       37/1    1.16824    1.16821 +/- 0.00485
+       38/1    1.18299    1.16874 +/- 0.00471
+       39/1    1.21418    1.17031 +/- 0.00480
+       40/1    1.11167    1.16835 +/- 0.00504
+       41/1    1.11545    1.16665 +/- 0.00516
+       42/1    1.11114    1.16491 +/- 0.00529
+       43/1    1.14227    1.16423 +/- 0.00517
+       44/1    1.14104    1.16355 +/- 0.00506
+       45/1    1.16756    1.16366 +/- 0.00492
+       46/1    1.13065    1.16274 +/- 0.00487
+       47/1    1.11251    1.16139 +/- 0.00492
+       48/1    1.14731    1.16101 +/- 0.00481
+       49/1    1.16691    1.16117 +/- 0.00469
+       50/1    1.19679    1.16206 +/- 0.00465
+ Creating state point statepoint.50.h5...
+
+ ===========================================================================
+ ======================&gt;     SIMULATION FINISHED     &lt;======================
+ ===========================================================================
+
+
+ =======================&gt;     TIMING STATISTICS     &lt;=======================
+
+ Total time for initialization     =  4.1700E-01 seconds
+   Reading cross sections          =  8.9000E-02 seconds
+ Total time in simulation          =  1.4728E+01 seconds
+   Time in transport only          =  1.4712E+01 seconds
+   Time in inactive batches        =  1.7890E+00 seconds
+   Time in active batches          =  1.2939E+01 seconds
+   Time synchronizing fission bank =  5.0000E-03 seconds
+     Sampling source sites         =  3.0000E-03 seconds
+     SEND/RECV source sites        =  2.0000E-03 seconds
+   Time accumulating tallies       =  1.0000E-03 seconds
+ Total time for finalization       =  1.0000E-03 seconds
+ Total time elapsed                =  1.5155E+01 seconds
+ Calculation Rate (inactive)       =  13974.3 neutrons/second
+ Calculation Rate (active)         =  7728.57 neutrons/second
+
+ ============================&gt;     RESULTS     &lt;============================
+
+ k-effective (Collision)     =  1.16131 +/-  0.00453
+ k-effective (Track-length)  =  1.16206 +/-  0.00465
+ k-effective (Absorption)    =  1.16096 +/-  0.00364
+ Combined k-effective        =  1.16120 +/-  0.00325
+ Leakage Fraction            =  0.00000 +/-  0.00000
+</pre></div>
+</div>
+<div class="highlight-python"><div class="highlight"><pre><span class="mi">0</span>
+</pre></div>
+</div>
+</div>
+<div class="section" id="tally-data-processing">
+<h1>Tally Data Processing<a class="headerlink" href="#tally-data-processing" title="Permalink to this headline">¶</a></h1>
+<p>Our simulation ran successfully and created statepoint and summary
+output files. We begin our analysis by instantiating a <code class="docutils literal"><span class="pre">StatePoint</span></code>
+object.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Load the last statepoint file</span>
+<span class="n">sp</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">StatePoint</span><span class="p">(</span><span class="s">&#39;statepoint.50.h5&#39;</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>In addition to the statepoint file, our simulation also created a
+summary file which encapsulates information about the materials and
+geometry. This is necessary for the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> module to properly
+process the tally data. We first create a <code class="docutils literal"><span class="pre">Summary</span></code> object and link it
+with the statepoint.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Load the summary file and link it with the statepoint</span>
+<span class="n">su</span> <span class="o">=</span> <span class="n">openmc</span><span class="o">.</span><span class="n">Summary</span><span class="p">(</span><span class="s">&#39;summary.h5&#39;</span><span class="p">)</span>
+<span class="n">sp</span><span class="o">.</span><span class="n">link_with_summary</span><span class="p">(</span><span class="n">su</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>The statepoint is now ready to be analyzed by our multi-group cross
+sections. We simply have to load the tallies from the <code class="docutils literal"><span class="pre">StatePoint</span></code>
+into each object as follows and our <code class="docutils literal"><span class="pre">MGXS</span></code> objects will compute the
+cross sections for us under-the-hood.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Load the tallies from the statepoint into each MGXS object</span>
+<span class="n">total</span><span class="o">.</span><span class="n">load_from_statepoint</span><span class="p">(</span><span class="n">sp</span><span class="p">)</span>
+<span class="n">absorption</span><span class="o">.</span><span class="n">load_from_statepoint</span><span class="p">(</span><span class="n">sp</span><span class="p">)</span>
+<span class="n">scattering</span><span class="o">.</span><span class="n">load_from_statepoint</span><span class="p">(</span><span class="n">sp</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>Voila! Our multi-group cross sections are now ready to rock &#8216;n roll!</p>
+</div>
+<div class="section" id="extracting-and-storing-mgxs-data">
+<h1>Extracting and Storing MGXS Data<a class="headerlink" href="#extracting-and-storing-mgxs-data" title="Permalink to this headline">¶</a></h1>
+<p>Let&#8217;s first inspect our total cross section by printing it to the
+screen.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="n">total</span><span class="o">.</span><span class="n">print_xs</span><span class="p">()</span>
+</pre></div>
+</div>
+<div class="highlight-python"><div class="highlight"><pre>Multi-Group XS
+    Reaction Type  =        total
+    Domain Type    =        cell
+    Domain ID      =        1
+    Cross Sections [cm^-1]:
+            Group 1 [6.25e-07   - 20.0      MeV]:   6.81e-01 +/- 1.88e-01%
+            Group 2 [0.0        - 6.25e-07  MeV]:   1.40e+00 +/- 5.91e-01%
+</pre></div>
+</div>
+<p>Since the <code class="docutils literal"><span class="pre">openmc.mgxs</span></code> module uses <a class="reference external" href="https://mit-crpg.github.io/openmc/pythonapi/examples/tally-arithmetic.html">tally
+arithmetic</a>
+under-the-hood, the cross section is stored as a &#8220;derived&#8221; <code class="docutils literal"><span class="pre">Tally</span></code>
+object. This means that it can be queried and manipulated using all of
+the same methods supported for the <code class="docutils literal"><span class="pre">Tally</span></code> class in the OpenMC Python
+API. For example, we can construct a
+<a class="reference external" href="http://pandas.pydata.org/">Pandas</a> <code class="docutils literal"><span class="pre">DataFrame</span></code> of the multi-group
+cross section data.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="n">df</span> <span class="o">=</span> <span class="n">scattering</span><span class="o">.</span><span class="n">get_pandas_dataframe</span><span class="p">()</span>
+<span class="n">df</span><span class="o">.</span><span class="n">head</span><span class="p">(</span><span class="mi">10</span><span class="p">)</span>
+</pre></div>
+</div>
+<div>
+<table border="1" class="dataframe">
+  <thead>
+    <tr style="text-align: right;">
+      <th></th>
+      <th>cell</th>
+      <th>group in</th>
+      <th>nuclide</th>
+      <th>mean</th>
+      <th>std. dev.</th>
+    </tr>
+  </thead>
+  <tbody>
+    <tr>
+      <th>1</th>
+      <td>1</td>
+      <td>1</td>
+      <td>total</td>
+      <td>0.668323</td>
+      <td>0.001264</td>
+    </tr>
+    <tr>
+      <th>0</th>
+      <td>1</td>
+      <td>2</td>
+      <td>total</td>
+      <td>1.293258</td>
+      <td>0.007624</td>
+    </tr>
+  </tbody>
+</table>
+</div><p>Each multi-group cross section object can be easily exported to a
+variety of file formats, including CSV, Excel, and LaTeX for storage or
+data processing.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="n">absorption</span><span class="o">.</span><span class="n">export_xs_data</span><span class="p">(</span><span class="n">filename</span><span class="o">=</span><span class="s">&#39;absorption-xs&#39;</span><span class="p">,</span> <span class="n">format</span><span class="o">=</span><span class="s">&#39;excel&#39;</span><span class="p">)</span>
+</pre></div>
+</div>
+<p>The following code snippet shows how to export all three <code class="docutils literal"><span class="pre">MGXS</span></code> to the
+same HDF5 binary data store.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="n">total</span><span class="o">.</span><span class="n">build_hdf5_store</span><span class="p">(</span><span class="n">filename</span><span class="o">=</span><span class="s">&#39;mgxs&#39;</span><span class="p">,</span> <span class="n">append</span><span class="o">=</span><span class="bp">True</span><span class="p">)</span>
+<span class="n">absorption</span><span class="o">.</span><span class="n">build_hdf5_store</span><span class="p">(</span><span class="n">filename</span><span class="o">=</span><span class="s">&#39;mgxs&#39;</span><span class="p">,</span> <span class="n">append</span><span class="o">=</span><span class="bp">True</span><span class="p">)</span>
+<span class="n">scattering</span><span class="o">.</span><span class="n">build_hdf5_store</span><span class="p">(</span><span class="n">filename</span><span class="o">=</span><span class="s">&#39;mgxs&#39;</span><span class="p">,</span> <span class="n">append</span><span class="o">=</span><span class="bp">True</span><span class="p">)</span>
+</pre></div>
+</div>
+</div>
+<div class="section" id="comparing-mgxs-with-tally-arithmetic">
+<h1>Comparing MGXS with Tally Arithmetic<a class="headerlink" href="#comparing-mgxs-with-tally-arithmetic" title="Permalink to this headline">¶</a></h1>
+<p>Finally, we illustrate how one can leverage OpenMC&#8217;s <a class="reference external" href="https://mit-crpg.github.io/openmc/pythonapi/examples/tally-arithmetic.html">tally
+arithmetic</a>
+data processing feature with <code class="docutils literal"><span class="pre">MGXS</span></code> objects. The <code class="docutils literal"><span class="pre">openmc.mgxs</span></code>
+module uses tally arithmetic to compute multi-group cross sections with
+automated uncertainty propagation. Each <code class="docutils literal"><span class="pre">MGXS</span></code> object includes an
+<code class="docutils literal"><span class="pre">xs_tally</span></code> attribute which is a &#8220;derived&#8221; <code class="docutils literal"><span class="pre">Tally</span></code> based on the
+tallies needed to compute the cross section type of interest. These
+derived tallies can be used in subsequent tally arithmetic operations.
+For example, we can use tally artithmetic to confirm that the
+<code class="docutils literal"><span class="pre">TotalXS</span></code> is equal to the sum of the <code class="docutils literal"><span class="pre">AbsorptionXS</span></code> and
+<code class="docutils literal"><span class="pre">ScatterXS</span></code> objects.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Use tally arithmetic to compute the difference between the total, absorption and scattering</span>
+<span class="n">difference</span> <span class="o">=</span> <span class="n">total</span><span class="o">.</span><span class="n">xs_tally</span> <span class="o">-</span> <span class="n">absorption</span><span class="o">.</span><span class="n">xs_tally</span> <span class="o">-</span> <span class="n">scattering</span><span class="o">.</span><span class="n">xs_tally</span>
+
+<span class="c"># The difference is a derived tally which can generate Pandas DataFrames for inspection</span>
+<span class="n">difference</span><span class="o">.</span><span class="n">get_pandas_dataframe</span><span class="p">()</span>
+</pre></div>
+</div>
+<div>
+<table border="1" class="dataframe">
+  <thead>
+    <tr style="text-align: right;">
+      <th></th>
+      <th>cell</th>
+      <th>energy [MeV]</th>
+      <th>nuclide</th>
+      <th>score</th>
+      <th>mean</th>
+      <th>std. dev.</th>
+    </tr>
+  </thead>
+  <tbody>
+    <tr>
+      <th>0</th>
+      <td>1</td>
+      <td>(0.0e+00 - 6.3e-07)</td>
+      <td>total</td>
+      <td>(((total / flux) - (absorption / flux)) - (sca...</td>
+      <td>4.884981e-15</td>
+      <td>0.011274</td>
+    </tr>
+    <tr>
+      <th>1</th>
+      <td>1</td>
+      <td>(6.3e-07 - 2.0e+01)</td>
+      <td>total</td>
+      <td>(((total / flux) - (absorption / flux)) - (sca...</td>
+      <td>1.221245e-15</td>
+      <td>0.001802</td>
+    </tr>
+  </tbody>
+</table>
+</div><p>Similarly, we can use tally arithmetic to compute the ratio of
+<code class="docutils literal"><span class="pre">AbsorptionXS</span></code> and <code class="docutils literal"><span class="pre">ScatterXS</span></code> to the <code class="docutils literal"><span class="pre">TotalXS</span></code>.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Use tally arithmetic to compute the absorption-to-total MGXS ratio</span>
+<span class="n">absorption_to_total</span> <span class="o">=</span> <span class="n">absorption</span><span class="o">.</span><span class="n">xs_tally</span> <span class="o">/</span> <span class="n">total</span><span class="o">.</span><span class="n">xs_tally</span>
+
+<span class="c"># The absorption-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection</span>
+<span class="n">absorption_to_total</span><span class="o">.</span><span class="n">get_pandas_dataframe</span><span class="p">()</span>
+</pre></div>
+</div>
+<div>
+<table border="1" class="dataframe">
+  <thead>
+    <tr style="text-align: right;">
+      <th></th>
+      <th>cell</th>
+      <th>energy [MeV]</th>
+      <th>nuclide</th>
+      <th>score</th>
+      <th>mean</th>
+      <th>std. dev.</th>
+    </tr>
+  </thead>
+  <tbody>
+    <tr>
+      <th>0</th>
+      <td>1</td>
+      <td>(0.0e+00 - 6.3e-07)</td>
+      <td>total</td>
+      <td>((absorption / flux) / (total / flux))</td>
+      <td>0.076219</td>
+      <td>0.000651</td>
+    </tr>
+    <tr>
+      <th>1</th>
+      <td>1</td>
+      <td>(6.3e-07 - 2.0e+01)</td>
+      <td>total</td>
+      <td>((absorption / flux) / (total / flux))</td>
+      <td>0.019319</td>
+      <td>0.000086</td>
+    </tr>
+  </tbody>
+</table>
+</div><div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Use tally arithmetic to compute the scattering-to-total MGXS ratio</span>
+<span class="n">scattering_to_total</span> <span class="o">=</span> <span class="n">scattering</span><span class="o">.</span><span class="n">xs_tally</span> <span class="o">/</span> <span class="n">total</span><span class="o">.</span><span class="n">xs_tally</span>
+
+<span class="c"># The scattering-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection</span>
+<span class="n">scattering_to_total</span><span class="o">.</span><span class="n">get_pandas_dataframe</span><span class="p">()</span>
+</pre></div>
+</div>
+<div>
+<table border="1" class="dataframe">
+  <thead>
+    <tr style="text-align: right;">
+      <th></th>
+      <th>cell</th>
+      <th>energy [MeV]</th>
+      <th>nuclide</th>
+      <th>score</th>
+      <th>mean</th>
+      <th>std. dev.</th>
+    </tr>
+  </thead>
+  <tbody>
+    <tr>
+      <th>0</th>
+      <td>1</td>
+      <td>(0.0e+00 - 6.3e-07)</td>
+      <td>total</td>
+      <td>((scatter / flux) / (total / flux))</td>
+      <td>0.923781</td>
+      <td>0.007714</td>
+    </tr>
+    <tr>
+      <th>1</th>
+      <td>1</td>
+      <td>(6.3e-07 - 2.0e+01)</td>
+      <td>total</td>
+      <td>((scatter / flux) / (total / flux))</td>
+      <td>0.980681</td>
+      <td>0.002617</td>
+    </tr>
+  </tbody>
+</table>
+</div><p>Lastly, we sum the derived scatter-to-total and absorption-to-total
+ratios to confirm that they sum to unity.</p>
+<div class="code python highlight-python"><div class="highlight"><pre><span class="c"># Use tally arithmetic to ensure that the absorption- and scattering-to-total MGXS ratios sum to unity</span>
+<span class="n">sum_ratio</span> <span class="o">=</span> <span class="n">absorption_to_total</span> <span class="o">+</span> <span class="n">scattering_to_total</span>
+
+<span class="c"># The scattering-to-total ratio is a derived tally which can generate Pandas DataFrames for inspection</span>
+<span class="n">sum_ratio</span><span class="o">.</span><span class="n">get_pandas_dataframe</span><span class="p">()</span>
+</pre></div>
+</div>
+<div>
+<table border="1" class="dataframe">
+  <thead>
+    <tr style="text-align: right;">
+      <th></th>
+      <th>cell</th>
+      <th>energy [MeV]</th>
+      <th>nuclide</th>
+      <th>score</th>
+      <th>mean</th>
+      <th>std. dev.</th>
+    </tr>
+  </thead>
+  <tbody>
+    <tr>
+      <th>0</th>
+      <td>1</td>
+      <td>(0.0e+00 - 6.3e-07)</td>
+      <td>total</td>
+      <td>(((absorption / flux) / (total / flux)) + ((sc...</td>
+      <td>1</td>
+      <td>0.007741</td>
+    </tr>
+    <tr>
+      <th>1</th>
+      <td>1</td>
+      <td>(6.3e-07 - 2.0e+01)</td>
+      <td>total</td>
+      <td>(((absorption / flux) / (total / flux)) + ((sc...</td>
+      <td>1</td>
+      <td>0.002619</td>
+    </tr>
+  </tbody>
+</table>
+</div></div>
+
+
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