setup initial isotopes concentrations

[josejavier]

Hello,

At present, I am performing decay calculations for spent fuel for long time periods (hundreds to thousands of years) using the compositions from the NEA benchmark on BUC-Phase VII, and I am using SERPENT2 as a decay code with the options 'decstep' or 'dectot' as coupled flux calculations are not needed.

The initial spent fuel composition contains isotopes for which there is no XS information, e.g. C-14, so SERPENT2 stops with the message like:

Nuclide 6014.03c not found in data libraries

I do not know if it is already possible, but it would be interesting to include the possibility of letting decay and track also this isotopes for which there is no XS information but that could be important for other purposes (public dose in the benchmark) and be present in an initial fuel composition. I would propose some indication in the material definition that the isotope should not be considered in the transport calculation, but only during the decay calculation. So the fix can be applied either in burnup or only decay calculations, therefore in any situation and will make SERPENT2 also a more complete decay code.

I suppose a way around is to create fake XS data files for those isotopes and the code probably will run also. Is that a possible partial solution with SERPENT2? Do you have another way around?

Thanks,
Jose

[Jaakko Leppänen]

Serpent actually tracks a lot of isotopes without cross sections, as they are produced in decay, fission and transmutation reactions. For some time now I have been trying to come up with a way to include these decay isotopes also in the material input, but it has turned to be quite complicated because the format of the mat card is already fixed. The main problem is that in order to form the decay and transmutation paths, the code needs a library id and temperature associated to each isotope, and these are defined only for nuclides with cross section data.

There are several users who have requested this capability in order to do restart calculations using material outputs printed from a previous run. For this purpose there is a special option "fix" for the material card. For example:

Code: Select all

mat fuel sum burn 1 fix 00d 300.0
C-14       1.7640e-09  
O-16       4.5260e-02  
Cl-36      1.0000e-06  
Ca-41      1.0000e-06  
Ni-59      1.0000e-06  
Se-79      4.7500e-07  
...

defines a material consisting of decay nuclides only (omitting the library id's in the isotope name means that the nuclide is handled as a decay nuclide without cross sections, the nuclides can also be defined using the ZAI, for example 60140 for C-14, 952421 for Am-242m, etc.). The two parameters after "fix" are the library id and temperature, which in decay calculation can be arbitrarily defined.

Could you try to see if this definition works with your application?

IMPORTANT NOTE: The fix option currently works differently in a decay and burnup calculation. If there are only decay intervals in the depletion history (decay calculation), the code forms automatically the decay paths and includes the necessary daughter isotopes. If there is one or more burnup intervals (burnup calculation), the code does not read any additional nuclides other than those listed in the material card. This means that if some important nuclide is missing in the composition, it is not included, and the decay and transmutation paths passing through it are cut short. This functionality is necessary to get the same list of nuclides included in the calculation if the user wants to continue a burnup calculation from a previous run in which the material compositions were written in the file. The list of nuclides included in the calculation is always printed in the <input>.out file.

[josejavier]

Thanks! It worked of course