Template and Runtime Files

As described in methods_coupling, SaltProc uses file-based coupling to iterface with depletion solvers. A user provides the template input file(s), and SaltProc makes a temporary copy used for running depletion steps that it can modify and update as needed.

Serpent 2 Template Files

A valid Serpent 2 template input file must have at least the following:

  • Burnable materials linked via the include card The burnable material cards must include the fix option.

  • Non-burnable materials instantiated directly in the file OR via an include card after the burnable materials.

  • Neutron particle parameters set via the set pop card.

  • Spontaneous fission yielf sublibray via the set sfylib card

Template input files should also have the following (they are not neccesary if the user has used the appropriate environment variables to specify them. See the links for each card for more details):

So a minimal example assuming the without the use of environment variables would look like:

% Cross section libraries
set acelib /home/user/xsdata/endfb71/endfb71.xsdata

%% SaltProc will write include card for geometry file here


% Sublibraries
set declib /home/user/xsdata/endfb71/endfb71.dec
set sfylib /home/user/xsdata/endfb71/endfb71.sfy
set nfylib /home/user/xsdata/endfb71/endfb71.nfy

% Include burnable materials
include /home/user/msbr_model/burnable_mats.ini

% Set neutron population
set pop 1000 40 10

And the material file containing burnable materials would look like:

mat uo2 -10.0 vol 0.554 burn 1 fix 82c 900
    8016.82c  1.0
    92234.82c 0.0004
    92235.82c 0.043
    92238.82c 0.9564
    92236.82c 0.0002

Users looking further cutomize the depletion step should read Serpent 2’s input syntax manual, in particular the set bumode card and the set pcc card.

OpenMC Template Files

Any valid Model object can be exported into a set of template files that will work with SaltProc. See the OpenMC docpages for more info.