Detailed modeling and massive tallying of nuclear reactors lead to memory overload for a single core processor.It could not be calculated by Monte Carlo particle transport with particle parallelism only.Domain decomposition is one of solutions.Domain decomposition needs to interchange particles between processors, so that inherit technique of pseudo-random number could not make identical results between serial and parallel.Two techniques of pseudo-random number are described to obtain identical results on different numbers of domains in a Monte Carlo particle simulation code.

In multi-step Monte Carlo simulation of time-dependent particle transport problems, particle-flag based physical quantity can be calculated by appropriate classification of diverse particle's attributes. Some particle-flag based physical quantities' fluctuation are strong since only very small fraction of total histories can make non-zero contribution and it is inefficient to deal with this problem by increasing purely total history number. A source bias algorithm is proposed to decrease stochastic error of target quantity by increasing number of source particle with a specific type only. Meanwhile, precision of non-target quantities are hardly decreased. A one-dimensional multi-layer model is utilized to display effect of the method.

In order to develop large-scale transport simulation (such as simulation of whole reactor core pin-by-pin problem), we developed a neutron photon coupled transport code JMCT. In this article, developing idea of a auto modeling tool JLAMT based on field oriented development is introduced. JLAMT developed several quick and assembly modeling tools. Data structure based on hierarchical geometry tree was designed. Automatic conversion and generation of physical model input file for GDML file format are made. By using those modeling tools, complex devices (include DAYAWAN whole-core model) were created, and transformed file was delivered to JMCT for transport calculation. Results were validated for correctness of visual modeling tools.

Multigroup neutron adjoint transport and automatic source particle biasing sampling function of JMCT is introduced. A press water reactor shielding neutron flux is achieved by JMCT with source biasing sampling technique, which is well accord with experiment. Its computation efficiency is higher than that of MCNP using geometry important method.

Mesh tally function of monte Carlo method can give a detailed and intensive calculation of flux distribution in specific volumes. To realize such function in JMCT mesh tally function are designed and realized. It supports non-uniform mesh in three kinds of orthogonal geometry (xyz of rectangular coordinates, rθz of cylindrical coordinates, and rθφ of spherical coordinates). Algorithm for rectangular coordinates is discussed. Calculation on DAYAWAN reactor core pin-by-pin model, Venus benchmark model and a 1-D ITER model verifies preliminarily correctness of JMCT mesh tally. Furthermore, U-array benchmark model is used to test serial performance of JMCT mesh tally. Both JMCT and MCNP5 use same xyz mesh grids and run under same condition. It shows that JMCT takes less time consuming and has higher performance dealing with xyz geometry.