Hello, everyone!
I want to use point reactor kinetics equations for calculating transient problems. The delayed neutron lifetime, fraction can be obtained by mgxs class. However, I can’t find a way to process the neutron generation time or prompt neutron lifetime. Does anyone know it,very gratefully!
Hello, although I’m not super fresh on the topic, I did my dissertation on transient problems in OpenMC. If you see this paper, equation 10 shows one way to calculate prompt neutron lifetime. Please see the PDF version of the paper, as the online version distorts equation 10.
You might also find it useful to poke around this branch of OpenMC on my GitHub for more.
Thank you for answering this question. That means that the adjoint flux is required for calculation neutron prompt life. It seems that OpenMC doesn’t have that function at the moment. I will try to generate the dynamic parameters using deterministic code. Thank you very much!
Hi again,
In the Github link I mentioned, an example of prompt neutron lifetime being calculated is shown in state.py in a function called pnl().
That refers to an adjoint flux which is calculated using a source iteration with matrices of Monte Carlo tallied entries. The code for that eigenvalue calculation can be found in aux_functions() in a function called compute_eigenvalue(). To obtain an adjoint, call the function with the transpose of the matrices.