Create Material Input Files

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OpenMOC is a multi-energy group method of characteristics solver. The material XML input file for OpenMOC must contain all of the nuclear data for each energy group for your simulation. It is important to note that every material must have the same number of energy groups. In addition, the energy groups are assumed to be in the standard order from highest to lowest energy.

Let us look at a simple example for a materials input file containing a single material with data for two energy groups:

<?xml version="1.0"?>
<materials>

<!-- Definition of materials based on 2 group data from "Fundamentals
       of Nuclear Reactor Physics" by Eugene Lewis
-->

  <material>
     <id>1</id>
     <energy>2</energy>
     <sigma_a>0.0038 0.184</sigma_a>
     <sigma_t>0.2208 1.604</sigma_t>
     <nu_sigma_f>0.0015 0.325</nu_sigma_f>
     <chi>1.0 0.0</chi>
     <sigma_s>0.1 0.117
	      0.0 1.42</sigma_s>

  />

</materials>

At the beginning of this file is a comment, denoted by a tag starting with . Comments, as well as any other type of input, may span multiple lines. One convenient feature of the XML input format is that sub-elements of the cell and surface elements can also be equivalently expressed of attributes of the original element, e.g. the materials file above could be written as:

<?xml version="1.0"?>
<materials>

<!-- Definition of materials based on 2 group data from "Fundamentals
       of Nuclear Reactor Physics" by Eugene Lewis
-->

  <material id="1" energy="2"
        sigma_a="0.0038 0.184"
	sigma_t="0.2208 1.604"
        nu_sigma_f="0.0015 0.325"
	chi="1.0 0.0"
	sigma_s="0.1 0.117
	         0.0 1.42"

  />

</materials>

###Materials Each material element must have all of the following attributes or sub-elements:

• id: A unique integer that can be used to identify the surface. Note: all ids must be less than 10,000.

• energy: The number of energy groups. Note: each material must have the same number of energy groups.

• sigma_a: The absorption cross-sections for each energy group in cm^-1.

• sigma_t: The total cross-sections for each energy group in cm^-1.

• nu_sigma_f: The fission cross-sections multipled by the average number of fission neutrons in cm^-1.

• chi: The fission spectrum for each energy group.

• sigma_s: The scattering matrix for all energy groups in cm^-1.