Discussion forum for Serpent users

The src card provides just the source, or in the case of a criticality source simulation, the initial guess for the fission source. The cross sections that you are calculating are averaged over the flux spectrum, which also depends on the system that you are modeling, the position of the activated m...

The activation step mode is designed to use cross sections from the previous interval. If that interval is a decay interval, the cross sections are not calculated. Unfortunately the routine cannot remember the values from the beginning of the calculation. Maybe one way around this would be to replac...

The results depend on the spectrum. Is the spectrum in your system comparable to the one where the reference value is given?

This could be related to the source sampling efficiency. If most of your source is in a region where the importance is low, the routine rejects most of the sampled points.

Could you post the full input (including the geometry and material definitions)?

There is the actstep option in the dep card that does just this.

My guess is, that there is some geometry error, possibly an overlap, related to material Am. If you replace "dm Am" with another material (e.g. "dm graphite") the results are consistent (TotalXS gives the same result as TotalRR/Totalflux).

Can you post the full input or send it o me by email?

To be precise, material density is used as a multiplier only for macroscopic reaction rates (negative response numbers). For microscopic rates the density in the material card plays no role.

All cell flux detectors are integrated over volume, so the flux detector gives you the integral flux, reaction rate detector the integral reaction rate. So if you want your flux to be in units n/cm²s, you need to divide by volume, etc. For the calculation of microscopic cross sections this is not ne...

At the moment there is no option to calculate sums of reaction rates. The cross sections given in the ACE data dictates what is available for the detectors.