Flux field calculation in time-dependent problems with Monte Carlo method and associated particles transportation equation are described.The random walk principle of time station sampling(TSS) method in dynamic system is introduced.A time station importance sampling(TSIS) method is developed and applied to decrease statistical fluctuation in global flux field calculation.Alternative importance function and its unbiased weight of modification are investigated.Numerical results and analyses on methods in specific model are given.The time station importance sampling(TSIS) biased method can be used to control statistical variance of global fine mesh flux adaptively in a same level.

A global variance reduction method is described.The method is based on a modified transport problem that can be solved by analog Monte Carlo with zero variance.It is developed by breaking up the whole into parts and coupling Monte Carlo method with discrete ordinates method and using approximate importance function to guide Monte Carlo simulation.Performance of the method for a one-dimensional sphere transport problem is demonstrated.The method is shown to produce lower variances globally compared to analog Monte Carlo.

Monte-Carlo method for simulationg the responses of a dual-space neutron porosity oil well-logging tool is developed and presented.For practical computations,the specific Monte-Carlo code introduced here uses implicit capture、weight and energy cut off、geometric splitting and Russian roulette and directional biasing techniques as primary variance reduction methods.Some of the techniques have been improved partly.Numerical tests of practical examples are discussed and analyzed,which confirmed the effectiveness of these techniques.

In this parper, the neutronics performance of a solid breeder blanket of a elementary designed fusion-fission hybrid experimental reactor has been studied in one-, two-, three-dimensional analysis and comparison. For being easy to describe the complicated geometry of the Tokamak torus, the neutron and gamma ray coupled transport calculation has been performed with the help of the Monte Carlo code MCNP. For avoiding uncertainties and errors involved in the generation and application of group constants and self-shielding effects, a continuous energy representation of the nuclear data base has been used.The main concern of the study has been the assessment of the spacial distribution of the tritium breeding ratio, the plutonium breeding ratio, and the power density in multi-dimensional configuration and the comparison between One-, two-, three-dimensional global calculation. The role and the limitation of one dimensional calculation has been pointed out.