I’m new on this forum but i hope that someone can help me. I want to compare Geant4 and MCNP results in a simulation that uses a thermal neutron source. I made a PhysicsList where I include the thermal scattering (it inherits QGSP_BIC_HP) and i read on the manual that to use the correct cross sections i have to call the elements using the TS prefix. But there are only around ten TS material (G4NeutronHPThermalScatteringNames.cc), for example i have to use B into the detector, but it is not included. Is there a way to add a new element with its cross sections? If yes, how can i do? Can i take the cross sections from ENDF?
I’m trying to use
void G4NeutronHPThermalScatteringData::BuildPhysicsTable(const G4ParticleDefinition &)
but It doesn’t seems to work.
Thank you for the attention.
P.S. i use Geant4 10.04