Search found 2304 matches

by Jaakko Leppänen
Tue Sep 15, 2020 10:06 am
Forum: Users
Topic: resonance integral calculation
Replies: 4
Views: 99

Re: resonance integral calculation

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...
by Jaakko Leppänen
Fri Sep 11, 2020 6:03 pm
Forum: Users
Topic: Inhibit neutron transport calculation
Replies: 3
Views: 99

Re: Inhibit neutron transport calculation

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...
by Jaakko Leppänen
Fri Sep 11, 2020 5:59 pm
Forum: Users
Topic: resonance integral calculation
Replies: 4
Views: 99

Re: resonance integral calculation

The results depend on the spectrum. Is the spectrum in your system comparable to the one where the reference value is given?
by Jaakko Leppänen
Wed Sep 09, 2020 12:19 pm
Forum: Users
Topic: How to get external source?
Replies: 35
Views: 17879

Re: How to get external source?

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)?
by Jaakko Leppänen
Wed Sep 09, 2020 10:44 am
Forum: Users
Topic: Inhibit neutron transport calculation
Replies: 3
Views: 99

Re: Inhibit neutron transport calculation

There is the actstep option in the dep card that does just this.
by Jaakko Leppänen
Mon Sep 07, 2020 12:26 pm
Forum: General
Topic: Reaction rate vs Burnup.
Replies: 8
Views: 193

Re: Reaction rate vs Burnup.

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).
by Jaakko Leppänen
Mon Sep 07, 2020 8:34 am
Forum: General
Topic: Reaction rate vs Burnup.
Replies: 8
Views: 193

Re: Reaction rate vs Burnup.

Can you post the full input or send it o me by email?
by Jaakko Leppänen
Sun Sep 06, 2020 2:00 pm
Forum: General
Topic: Reaction rate vs Burnup.
Replies: 8
Views: 193

Re: Reaction rate vs Burnup.

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.
by Jaakko Leppänen
Sun Sep 06, 2020 10:26 am
Forum: General
Topic: Reaction rate vs Burnup.
Replies: 8
Views: 193

Re: Reaction rate vs Burnup.

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...
by Jaakko Leppänen
Fri Sep 04, 2020 5:02 pm
Forum: Users
Topic: Print 1-group XS with xsplot
Replies: 4
Views: 82

Re: Print 1-group XS with xsplot

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.