Hi Jaakko,
The scattering matrix generated by the detector is an "averaged" homogenized crosssection for the whole model, is that correct? e.g. INF_SCATT0, INF_SCATTP0 ... etc.
Is there a way to generate material dependent macroscopic crosssection, including total, absorption, nu*fission, and P0, P1, P2, P3 .... scattering matrices?
I am trying to generate some multigroup crosssection for transport code calculations.
Thanks!
Vince
Crosssection generation questions

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Re: Crosssection generation questions
Hi Vince,
Serpent does the group constant generation on universe basis, which means that you can get materialwise homogenized group constants by specifying different materials as different universes. The most simple approach would probably be to create an inf surface and create separate infinite cells/universes for different materials. Then you can replace the materials in cell, pin etc. definitions by filling in the corresponding universe instead.
Finally, set group constant generation on for all relevant universes using the set gcu card.
Ville
Serpent does the group constant generation on universe basis, which means that you can get materialwise homogenized group constants by specifying different materials as different universes. The most simple approach would probably be to create an inf surface and create separate infinite cells/universes for different materials. Then you can replace the materials in cell, pin etc. definitions by filling in the corresponding universe instead.
Finally, set group constant generation on for all relevant universes using the set gcu card.
Ville
Re: Crosssection generation questions
Sorry for somehow repetitive and basic question, however as I am not sure if I fully understood the answers, I'd like to ask again.
I would like to estimate the effective macroscopic cross section of a medium with few isotopes that have high microscopic thermal absorption cross section. I have an external neutron source, normalised to its strength (set src rate). All cells are 2D cylinders, I use black boundary condition (set bc 1)
Is it enough to define a detector like the following?
% flux detector in the material in 1 energy group, m_1 is the cell with my material, u03 the universe I used as well in set gcu u03
det Flux_1 de 1 dc m_1 du u03
%macroscopic reaction rate / flux
det Sigma_1 de 1 dc m_1 du u03 dr 1 void dt 3 Flux_1
When I compare the results with my total macroscopic cross section (Sigma) I calculate with transmission (for the same material, source and thickness  I1=I0exp(Sigma*d) ), I get very different results and I can't figure out what am I doing incorrectly...
Thank you in advance for any help!
I would like to estimate the effective macroscopic cross section of a medium with few isotopes that have high microscopic thermal absorption cross section. I have an external neutron source, normalised to its strength (set src rate). All cells are 2D cylinders, I use black boundary condition (set bc 1)
Is it enough to define a detector like the following?
% flux detector in the material in 1 energy group, m_1 is the cell with my material, u03 the universe I used as well in set gcu u03
det Flux_1 de 1 dc m_1 du u03
%macroscopic reaction rate / flux
det Sigma_1 de 1 dc m_1 du u03 dr 1 void dt 3 Flux_1
When I compare the results with my total macroscopic cross section (Sigma) I calculate with transmission (for the same material, source and thickness  I1=I0exp(Sigma*d) ), I get very different results and I can't figure out what am I doing incorrectly...
Thank you in advance for any help!

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Re: Crosssection generation questions
How do you calculate the total macroscopic XS? Off the top of my head, some stuff to consider:
 It could use gcu option or detectorbased for group constant generation.
 If choose 'gcu' option, keep in mind that group constant generation is calculated on universe basis. To obtain materialwise group constant define cell/universes for each material; If multiple materials are define in the same universe (i.e. whole geometry, universe 0), the result will be homogenised group constant integrated over the universe/geometry.
 If choose to output macroscopic XS through detectors, there is no need to keep 'gcu' option active (set gcu 1)
 'dm'/'dc'/'du' describes the spatial domain where the detector is scored (region over which the response is integrated).  If you have defined a single material/cell/universe, the definition seems redundant.
 'void' option in the detector card means the response is not preassigned with a specific material (when the detector scored in a collision, the XS is taken from the material at the collision point  allowing calculating integral reaction rates over regions composed of multiple materials); if it is not the case, define material/nuclide and substitute 'void'.
 the results of detector are integrated values without division by volume; including 'dv' card with volume to consider it.
 It could use gcu option or detectorbased for group constant generation.
 If choose 'gcu' option, keep in mind that group constant generation is calculated on universe basis. To obtain materialwise group constant define cell/universes for each material; If multiple materials are define in the same universe (i.e. whole geometry, universe 0), the result will be homogenised group constant integrated over the universe/geometry.
 If choose to output macroscopic XS through detectors, there is no need to keep 'gcu' option active (set gcu 1)
 'dm'/'dc'/'du' describes the spatial domain where the detector is scored (region over which the response is integrated).  If you have defined a single material/cell/universe, the definition seems redundant.
 'void' option in the detector card means the response is not preassigned with a specific material (when the detector scored in a collision, the XS is taken from the material at the collision point  allowing calculating integral reaction rates over regions composed of multiple materials); if it is not the case, define material/nuclide and substitute 'void'.
 the results of detector are integrated values without division by volume; including 'dv' card with volume to consider it.
 Ana
Re: Crosssection generation questions
Dear Ana,
First of all thank you a lot for your reply.
What I am trying to check if I can usethe BeerLambert law to estimate macroscopic cross section of absorber in material. I want to compare it with macroscopic cross section that I simply tallied with det option (Macroscopic reaction rate / flux) and calculated by hand ( microscopic cross section*at. density).
BeerLambert (Sigma = ln(I0/I1)/x) gives me too high values and I am trying to understand if perhaps I defined something incorrectly...
First of all thank you a lot for your reply.
What I am trying to check if I can usethe BeerLambert law to estimate macroscopic cross section of absorber in material. I want to compare it with macroscopic cross section that I simply tallied with det option (Macroscopic reaction rate / flux) and calculated by hand ( microscopic cross section*at. density).
BeerLambert (Sigma = ln(I0/I1)/x) gives me too high values and I am trying to understand if perhaps I defined something incorrectly...
 Jaakko Leppänen
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Re: Crosssection generation questions
What kind of medium and source do you have? How do you calculate the intensity?
 Jaakko