Update 1.1.11

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Jaakko Leppänen
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Update 1.1.11

Post by Jaakko Leppänen » Thu May 20, 2010 11:20 am

Update to Serpent version 1.1.11 was sent to registered users by e-mail on May 20, 2010. If you are using the code and didn't receive the update, send me e-mail and I'll send you the file and add you to the mailing list.

The major new thing in version 1.1.11 is the implementation of an external (fixed) source calculation mode. For the user, this new capability means another input structure that defines the source distribution. I started writing the description in the User's Manual (Chapter 9). The current definition allows simple point sources and volumetric sources in cells and materials. Source energy can be fixed to a single value, or sampled from a distribution in the ACE format data. Angular distribution is either isotropic or monodirectional. In future versions I intend to add at least the capability to use tabulated distributions read from separate input files. Other suggestions are also welcome. The external source definition can also be used as the initial guess for a criticality source simulation.

As I validated the new calculation routines, I was able to check for the first time how the interaction physics work for a single, non-fissile nuclide at any given energy. I did discover some inconsistensies compared to MCNP results, mainly at higher (14 MeV) energies where previous validations have not been carried out. Some problems were fixed, but some still remain. It should be noted, however, that most of the discovered problems occur at energies well beyond the emission spectrum of fission neutrons, so they should not be reflected in criticality source calculations. I'll keep debugging the reaction laws, but for the moment please be cautious when using the new calculation mode, especially for high-energy neutrons.

Another improvement was made for the micro-group spectrum method used to speed up the calculation of one-group transmutation cross sections for burnup calculation. The old implementation did not account for probability table sampling in the unresolved resonance region, which caused problems with fast reactor applications. This flaw is now fixed, and unresolved resonance cross sections are handled separately.

Minor changes include user-defined material colors in geometry plots and the capability to define the range of actinide mass chains included in burnup calculation.

I have rerun the infinite-lattice test calculations, and the results can be found at the website. The small flux anomaly in the CANDU case for JEF-2.2 (see related topic) is now corrected, but two new problems have emerged (I have probably broken something while fixing other reaction laws). Fission nubar in energy group 3 in the SFR case with JEF-2.2 (ures) is too high. Fast (group 1) capture cross section for the prismatic HTGR calculation using ENDF/B-VI.8 (no ures) is also slightly over-predicted. The differences are small but statistically significant. I'll keep looking for the cause.

Unfortunately I did not find the time to look into the random number generator issues, briefly discussed at the forum, but I'll try to do it for the next update.
- Jaakko

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