Add a discussion of different tools and analyses for radiation shielding.

.gitignore

@@ -0,0 +1 @@
+gh-build

source/_themes/uwtheme/layout.html

@@ -15,6 +15,7 @@
       <ul id="menu-main" class="sf-menu">
         {%- block rootrellink %}
         <li class="menu-item"><a href="{{ pathto(master_doc) }}">{{ shorttitle|e }}</a></li>
+        <li class="menu-item"><a href="DAGMC">DAGMC</a></li>
         {%- endblock %}
         {%- for rellink in rellinks|reverse %}
         <li class="menu-item" {% if loop.first %}style="margin-right: 10px"{% endif %}>
@@ -33,7 +34,7 @@
 {%- macro uwsidebar() %}
 <div id="sidebar">
  <div id="sidebar_section"></div>
- <div id="sidebar_section">{{ toctree() }} </div>
+ <div id="sidebar_section">{{ toc }} </div>
 </div>
 {%- endmacro %}
 

source/_themes/uwtheme/static/default.css_t

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-.pre { font-family: courier; font-size: 90%;}
 
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@@ -90,14 +88,17 @@ strong { font-weight: bold; }
 em { font-style: italic; }
 li h3, li p, li a { font-size: 1em; }
 
+.headerlink { color: #ffffff; }
+.headerlink:hover { color: #b70101; }
+
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   color: #000;
   background: #FFFFFF url(../_static/images/background.jpg) repeat-x top;
 }
 
 body {
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+  width: 1260px;
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   font-family:georgia,Times New Roman,serif;
   font-size: 2em;
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-  top: 26px;
+  top: 16px;
   left: 87px;
-  width: 343px;
-  height: 34px;
+  width: 500px;
+  height: 72px;
   overflow: hidden;
 }
 .home #siteTitle {
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 /*layout*/
 
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+  width: 1258px;
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 }
 .page #content {
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-  width: 640px;
+  width: 940px;
   padding-top: 40px;
 }
 .page #sidebar {
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   width: 250px;
 }
 
+.align-right {
+  float: right;
+  padding-left: 10px;
+}
+
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+  position: fixed;
+  width: 245px;
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+#sidebar ul, ol { list-style: none; }
+
+#sidebar ul ul { margin: 0 0 0 1em; list-style-type: circle; }
+#sidebar ul ul ul { margin: 0 0 0 1em; list-style-type: square; }
+
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 .page #sidebar .related_posts {
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+
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+
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+.pre {font-family:Courier; background: #ffefef; border: 1px solid #ffaaaa;}
 
+dl.docutils dt {
+   text-decoration: underline;
+   margin: 1em 0 0 0;
+}
+dl.docutils dd {
+   margin: 0 0 0 1em;
+}
 /*Innovations single*/
 
 .single .post .label {
@@ -750,9 +776,10 @@ table {
 margin-top: 16px;
 margin-bottom: 2em;
 text-align: left;
-border: 1px solid #666;
+border: 0px solid #666;
 border-collapse: collapse;
 font-size: .9em;
+width: 100%;
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 caption {
 text-align: left;

source/index.rst

@@ -7,10 +7,10 @@ the same way, fusion shielding is necessary to protect the system's
 components from the radiation produced by the fusion system they are
 supporting.
 
-This project includes a variety of tools, methodologies and workflows
-that are relevant for the nuclear analysis of complex system,
-including fusion power plant designs, accelerator irradiation
-facilities and critical reactor systems.
+This project includes a variety of `tools, methodologies and workflows
+<map.html>`_ that are relevant for the nuclear analysis of complex system,
+including fusion power plant designs, accelerator irradiation facilities and
+critical reactor systems.
 
 Single Physics Analysis Tools
 ;;;;;;;;;;;;;;;;;;;;;;;;;;;;;

source/map.rst

@@ -0,0 +1,245 @@
+Concepts and Tools for Radiation Shielding Analysis
+====================================================
+
+Over recent years, the set of concepts, methodologies and tools that are
+available for the analysis of radiation transport and effects in shielding
+systems has grown dramatically.  Here we attempt to describe the map of these
+capabilities with mention of specific tools that have gained, or are gaining,
+broader application.
+
+Fundamental Concepts
+--------------------
+
+The fundamental analysis need for any radiation shielding problem is
+**radiation transport**, for which two primary approaches are in widespread
+use: Monte Carlo radiation transport and discrete ordinates radiation
+transport.  Many of the effects of radiation can be calculated directly by the
+tools that implement these capabilities, but determining the induced
+radioactivity caused by neutron irradiation requires **activation analysis**
+to be used in conjunction with radiation transport.
+
+Monte Carlo Radiation Transport
+++++++++++++++++++++++++++++++++
+
+Monte Carlo radiation transport directly simulated the stochastic behavior of
+particles that leave a source and travel around the engineered system.  If
+enough particles are simulated using independent sequences of pseudo-random
+numbers, then their mean behavior can be taken as representative of the real
+system.  The single biggest advantage of Monte Carlo radiation transport is
+its ability to model the entire phase space of radiation transport (space,
+direction and energy) as continuous.  Continuous modeling of space allows for
+arbitrarily complex geometric models with need for approximation.  Taken
+together with continuous treatment of direction, this also means that
+following particles through complex streaming paths is also possible with
+these methods.  Finally, continuous treatment of energy allows less
+approximation of resonance behaviors for particles that are slowing down
+through that energy region.  This continuous modeling capability is the
+primary reason that Monte Carlo radiation transport is widespread for this
+kind of analysis.
+
+The biggest drawback of Monte Carlo techniques is in the modeling of deep
+penetration problems with substantial shielding.  Since such systems are
+designed to prevent most particles from reaching the far side of such shields,
+and Monte Carlo radiation transport essentially simulates the behavior of
+these particles, then very few simulated particles will penetrate the shield.
+This, in turn, results in very low quality results behind thick shields, since
+the quality of statistical results relies on a substantial number of
+contributions.  While it is possible to simply increase the total number of
+simulated particles, a number of schemes have been introduced, collectively
+known as *variance reduction* techniques to help improve this situation.
+
+The most widely used tool for Monte Carlo radiation transport is:
+
+* MCNP (Monte Carlo N-Particle), versions 5 and 6, written & maintained by Los
+  Alamos National Laboratory.
+
+While other tools exist, their user-base is dwarfed by that of MCNP: Fluka (CERN)
+Geant4 (CERN), Monaco (ORNL), SHIFT (ORNL), Tripoli (CEA).
+
+Discrete Ordinates Radiation Transport
++++++++++++++++++++++++++++++++++++++++
+
+While a variety of so-called deterministic methods exist, the discrete
+ordinates method is the most commonly used in radiation shielding analysis.
+This method discretizes space, direction and energy, and then solves a
+difference formulation of the radiation transport equation.  The most
+important advantage of deterministic methods is their ability to find global
+solutions with uniform quality.  Discretization in space introduces some
+approximations that can be largely overcome by refining the spatial
+resolution.  Discretization in direction, however, results in two related
+effects that can undermine the quality of results.  First, for transport in
+large (near) vacuum regions, particles are transported only along a discrete
+set of directions introducing so-called *ray effects*.  Furthermore, for
+arbitrarily oriented streaming paths and/or dog-legs, there may be no discrete
+angle that is aligned with such a path, inhibiting accurate streaming
+solutions.
+
+The biggest benefit of discrete ordinates techniques is that the quality of
+the solutions is independent of the thickness of the shielding region, and
+results can generally be found much more quickly than with Monte Carlo
+radiation transport.
+
+Use of discrete ordinates software is less wide-spread with less
+concentration in any single tool.  Some tools with user communities that
+extend beyond their developer communities include:
+
+* PARTISN (LANL)
+* DORT/TORT (ORNL)
+* Attila (Varian)
+* Exnihilo (ORNL)
+
+Activation Analysis
++++++++++++++++++++++
+
+Activation analysis generally combines the results of neutron radiation
+transport with cross-section data to determine the transmutation rates of all
+the nuclides in a material.  The resultant nuclides may also undergo
+transmutation or decay, and so on.  The equations that describe the evolution
+of the isotopic composition are a relatively simple set of first order ODEs,
+often referred to as the Bateman equations.  However, this is generally a very
+stiff set of equations, as governed by the transmutation and decay rates that
+have time constants that span from milliseconds to millions of years.  There a
+variety of mathematical and computational approaches to solving matrix
+exponential that arises from this set of equations.
+
+The majority of tools available for isotopic inventory calculations are
+derived from fission burnup codes, solving the inventory problem at a single
+point in space, that is, a single material with single neutron flux spectrum.
+They generally employ different computational approaches.  Some commonly used
+tools are ORIGEN (ORNL), FISPACT (CCFE), and CINDER (LANL).  ALARA (UW) is a
+special purpose activation tool that solves for multiple points in space
+simultaneously.
+
+
+Geometry Representations and CAD-based Analysis
+-------------------------------------------------
+
+Since the early 2000's, a number of different teams have pursued the ability
+to perform radiation transport and activation on geometries that are based on
+computer-aided designs (CAD).  For Monte Carlo radiation transport, the most
+common approach (Germany/KIT/McCAD, China/FDS/MCAM, USA/Raytheon/TopAct, etc)
+is to simplify the CAD models to include only the same geometric surfaces that
+can be represented in the constructive/combinatorial geometry representations
+that are native to the standard Monte Carlo tools, decompose the CAD regions
+into more simpler regions, and generate an input file in the standard format
+for those tools.  An alternative approach (USA/University of Wisconsin/DAGMC)
+is to modify the Monte Carlo software itself to perform its geometry queries
+directly on CAD-based representation of the geometry.  
+
+For discrete ordinates radiation transport, all tools but Attila rely on
+structured orthogonal grids for their geometry representation.  This requires
+some degree of approximation, either by choosing a dominant material for each
+voxel in that mesh or by mixing the materials in each voxel according to their
+volume fraction in that voxel.  CAD-based geometries can be queried according
+to either of these approaches to populate a structured grid representation of
+the geometry.
+
+Attila implements the discrete ordinates method using a finite element
+approach on a tetrahedral mesh, allowing the mesh to conform to arbitrarily
+complex shapes of the geometry and for each mesh element to contain a single
+material.  Attila is able to directly import CAD-based geometry in order to
+generate mesh for its transport algorithm.
+
+Frequently, the CAD models available for a given system have been generated
+without engineering analysis in mind.  As such, these models include details
+that are unnecessary for nuclear analysis as well as overlaps of different
+components.  The cleaning and repair of CAD-based geometries is currently not
+automatable and benefits greatly from experience working with previous models
+for this purpose.
+
+Automated Variance Reduction Techniques
+----------------------------------------
+
+As indicated above, Monte Carlo radiation transport can be challenged by
+problems that involve deep penetration of thick shields, a characteristic of
+most shielding problems.  A number of techniques, generally known as variance
+reduction techniques, exist to accelerate these calculations for a specific
+response.  However, as the problems become more complex, it becomes difficult
+to manually configure those techniques for maximum benefit.  Over the last
+15-20 years, a number of approaches have been developed to automate the
+configuration of variance reduction.  Some approaches (MCNP weight window
+generator, MAGIC) use iteration, where one round of Monte Carlo simulation is
+used to improve a guess for the variance reduction parameters for the
+successive round.
+
+An alternative approach is to use the results of a deterministic calculation
+to provide the parameters for variance reduction.  This is effective because
+the deterministic calculation is generally much faster than a Monte Carlo
+calculation with uniform quality of results.  ORNL has implemented the CADIS
+and FW-CADIS methodologies in the tool ADVANTG for this purpose, when used in
+conjunction with MCNP.  CADIS will optimize variance reduction parameters for
+a single response while FW-CADIS can be employed to optimize variance
+reduction parameters for many responses, including global solutions.
+
+Types of Analysis
+------------------
+
+These tools are used in various combinations to accomplish different types of
+analysis.  Depending on the combination of tools, different skills and
+expertise are necessary.
+
+Radiation transport, damage, heating, and tritium breeding
+++++++++++++++++++++++++++++++++++++++++++++++++++++++++++
+
+A basic radiation transport calculation can be configured to provide a variety
+of nuclear responses in addition to the neutron flux distribution.  These
+responses must be linear combinations of the neutron fluxes, typically based
+on cross-sections or similar nuclear data.  It is common to estimate radiation
+damage, in terms of displacements per atom as well as gas production, as well
+as heating from neutron and photon radiation.  In fusion systems, it is also
+of interest to estimate the number of tritium atoms that are produced in
+breeding regions.
+
+Increasingly, these analyses rely on CAD-based geometries to define the
+original geometry.  In addition, most of these analyses benefit from the use
+of automated variance reduction techniques.
+
+Basic expertise: 
+
+* Monte Carlo radiation transport, usually MCNP5/6.  
+
+Additional expertise: 
+
+* CAD-model preparation when using CAD-based approaches.
+* Automated variance reduction when necessary
+
+Shutdown Dose Rate Analysis
+++++++++++++++++++++++++++++
+
+Another important consideration in any system with neutrons is the impact of
+transmutation caused by those neutrons.  The most common consequence is the
+activation of structural components that results in a distributed photon
+source that dominates the radiation environment when the main system stops
+operating.  The shutdown dose rates that arise from these photons can be
+calculated throughout the facility.
+
+The most rigorous approach for this analysis is to first perform neutron
+transport to determine the spatial distribution of the multi-group neutron
+flux throughout the system.  At each point in space for which the neutron flux
+is known, an activation problem is performed to determine the photon source at
+that point.  Finally, the superposition of all photon sources is used for a
+photon transport problem in which the spatial distribution of the photon dose
+is determined.  Because of its separate transport steps for neutrons and
+photons, this approach is often referred to as the "rigorous 2-step approach"
+(R2S). For very large systems, it may be necessary to use spatial
+decomposition of the neutron flux results and photon source results.
+
+In most cases, a large mesh is used for both distributions of neutron flux and
+consequent photon sources, and perhaps a different mesh for the photon dose
+distribution.  This results in up to 1 million separate activation calculation
+points, and thus requires automation to be tractable.  An automated R2S
+approach has been implemented robustly as part of the PyNE toolkit, coupling
+MCNP to ALARA, with support for CAD-based geometries.
+
+Basic expertise: 
+
+* Monte Carlo radiation transport, usually MCNP5/6.  
+* Activation analysis using FISPACT, ALARA, or ORIGEN
+
+Additional expertise: 
+
+* CAD-model preparation when using CAD-based approaches.
+* Automated variance reduction may be used for the photon dose phase.
+
+Experimental work is underway (ORNL/UW) to determine how to best use automated
+variance reduction for the neutron transport step.