Running GROUPR

Setting up GROUPR runs can be daunting, because it has many possibilities for giving the user complete control over which reactions and data types are processed. However, for many applications, the user can use automatic modes that greatly simplify the input. Here is a simple example of a deck that computes infinitely dilute cross sections for carbon using a built-in group structure and weight function.
             20 21 0 22 / ENDF on tape20, PENDF on tape21, GENDF on tape22
             1306 3 3 3 3 1 1 1 / 30 groups, CLAW weight, P3
             'carbon from ENDF/B-V'/
             300 / one temperature
             1.e10 / infinite dilution only
             3/ process all cross sections on PENDF
             6/ process all matrices
             16/ process all photon production reactions
             0/ end of this temperature
             0/ end of groupr input

A clearer understanding of this input lines can be obtained by studying the GROUPR input specifications in the online version, a printed NJOY manual, or the files in the NJOY distribution. The notes after the "slash" terminator explain most of the features of this input deck.

Because not all user need all possible reactions, the automatic input process skips over thermal data (MT=221--250) and delayed neutron data (MT=455). If you need thermal data to be group averaged, include something like this in your deck:

             3/ all reactions but thermal
             3 229 'graphite inelastic'/
             3 230 'graphite elastic'/
             6/ all matrices but thermal
             6 229 'graphite inelastic'/
             6 230 'graphite elastic'/

Similarly, if you need delayed-neutron production and spectra, include something like this in your deck:
             3 455/ delayed nubar
             5 455/ delayed spectra

The code will automatically provide spectra for all the six time groups of delayed neutrons. As discussed in the previous subsection, it is the responsibility of a subsequent code to merge the delayed neutron data with the prompt data for the calculation of a proper fission source. As a final example, to obtain self-shielded cross sections and scattering matrices, you have to ask for several temperatures and several σ0 values. It is not necessary to reprocess all the reactions at the higher temperatures; just include the reactions that occur at resonance energies:
             20 21 0 22
             1050 3 0 3 3 2 3 1
             '94-pu-238 from ENDF/B-IV'/
             300 900
             1e10 1e4 1000 100 10 1
             3/ all reactions at 300K
             6/ all matrices at 300K
             0/ end of first temperature (300K)
             3 1 'total'/ only resonance reactions at 900K
             3 2 'elastic'/
             3 18 'fission'/
             3 102 'capture'/
             6 2 'elastic'/ yes, you can self shield the elastic matrix
             0/ end of second temperature (900K)
             0/ end of groupr deck

GROUPR has other capabilities, including custom group structures, custom weight functions, the flux calculator, charged particle transfer matrices, and activation cross section processing, that can be explored in the detailed documentation


23 January 2013 T-2 Nuclear Information Service