openmc-dev
Material with 1/v cross section for Adjoint weighted neutron generation time
Description
To calculate accurately the effective mean generation time $\Lambda_{eff}$, a technique alternative to the already implemented IFP is presented in [1]. It adds an artificial material with a $\sigma_{capt} \propto 1/v$ and solves the following equation: $\Lambda_{eff} = \frac{-\Delta\rho}{c}$ where $c$ is the amplitude of the cross section that has to be chosen by the user, according to the system under study.
This modification can be implemented outside OpenMC by creating this artificial material in the Nuclear Data Library used, remaining anyway the problem of apply the perturbation homogeneously along the whole system (in order to perturb the same amount in all the geometry).
Considering that it could be a very useful tool for Kinetic Parameters with less computational cost than the IFP. I would like to suggest the implementation of this method in order to avoid modification of Nuclear Data Libraries which can be validated for design/safety studies.
[1] Verboomen, B. and Haeck, W. and Baeten, P. "Monte Carlo calculation of the effective neutron generation time". Annals of nuclear energy vol 33, N. 10. pages 911--916. 2006.
