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

翻译：本文继续我们在[arXiv:1809.00827v2]、[arXiv:181.01779v4]和[arXiv:1901.00220v3]所做工作的基础上处理中子运输赤道(NTE),其中描述了中子通过异质裂变介质的通量,我们的目的是分析现有的和新的蒙特卡洛(MC)算法,目的是模拟与基本模型相关的铅密封值,这一数量在核监管工业中至关重要,NTE必须针对与核反应堆核心、辐照性医院设备、食品辐照设备等相适应的复杂无异领域加以解决。我们包括对此类MC算法的复杂分析,指出以前在文献中没有出现过这种承诺。新的MC算法提供了精准成本的各种利弊,以及更方便的计算平行的可能性。