Monte Carlo Methods for the Neutron Transport Equation

This paper continues our treatment of the Neutron Transport Equation (NTE) building on work in [arXiv:1809.00827v2], [arXiv:1810.01779v4] and [arXiv:1901.00220v3], which describes flux neutrons through inhomogeneous fissile medium. Our aim is to analyse existing novel Monte Carlo (MC) algorithms, aimed at simulating lead eigenvalue associated with underlying model. quantity principal importance nuclear regulatory industry for NTE must be solved complicated inhomogenous domains corresponding reactor cores, irradiative hospital equipment, food irradiation equipment so on. We include a complexity analysis such MC noting that no undertaking has previously appeared literature. The new algorithms offer variety advantages disadvantages accuracy vs cost, as well possibility more convenient computational parallelisation.