Discussion forum for Serpent users

Serpent 2.1.32 was released on February 24, 2021.

Bug reports etc. specifically related to update 2.1.32 under this thread.

Unfortunately, no. The transport simulation operates on microscopic level, and in addition to reaction probabilities (characterized by cross sections) the data includes energy and angular distributions that describe the scattering physics of individual neutron-target interactions. The connection to ...

I don't think this is related to the precision of the solution algorithm. The differences between the default CRAM and an analytical solution are very small. The solution method used for the Bateman equations should not be a major source of discrepancy when comparing the results of two burnup calcul...

I've also faced similar issues. ENDF/B-VII doesn't include the emission spectra for spontaneous fission neutrons.

Thank you for pointing this out. The makeburnmatrixmsr.c routine was separated from makeburnmatrix.c to avoid errors in the main burnup routine when the continuous reprocessing feature was added. It seems that some bug fixes applied later have not been carried out to the MSR routine. We'll add this ...

Try dividing all distributions with the mean value of the spectrum. Also, I think the energy group boundaries of the 33g Serpent spectrum may be shifted to the right by one group.

As Ana pointed out, some earlier MCNP versions had delayed neutron emission switched off by default in external source mode. This default behavior was also applied in Serpent.

A website for 2020 UGM has been set up at: http://montecarlo.vtt.fi/mtg/2020_Munich/index.htm

You can also try checking the consistency of the two models by calculating the volumes by Monte Carlo sampling. Serpent has the -checkvolumes option for that, and similar method can be used with MCNP by setting the materials to void (see MCNP manual for instructions). Since the discrepancies seem to...

The server was updated. The Wiki is up again.