Study On Deterministic Neutron Transport Calculation For Stochastic Medium Fuel

The stochastic medium fuel can be defined as that the fissile material is filled in a micro-spherical particle,and plentiful particles are dispersed randomly in the stochastic medium space within the matrix material.The novel type of fuel can prevent radioactive production from being released into the environment in a serious accident,which features inherent safety for the advanced reactor.However,the large quantity and random locations of particles bring challenges to the conventional core physics calculation.With the gradual increase in the requirement for the reactor safety features recently,it is necessary to solve the neutronics calculation difficulties of the accident-tolerant stochastic medium fuel.The focus of this thesis is the deterministic neutron transport calculation method of the stochastic medium fuel.For the special randomness feature of dispersed particles,the matrix chord length probability distribution is studied.The matrix chord length is the chord length between particles.The matrix chord length probability distribution describes the relative location among the dispersed particles.The matrix chord length probability distribution is obtained by sampling the matrix chord lengths on a large scale.The distribution functions in different sizes of stochastic medium and dispersed particles,different volumetric packing fraction of particles are analyzed.In the previous investigations,the Markovian distribution of the binary infinite stochastic medium is usually employed to describe the practical matrix chord length distribution.The effect on the core physics calculation caused by the Markovian distribution is evaluated in this thesis.The study on the matrix chord length distribution is the theoretical foundation for the Method of Characteristic(MOC)in the stochastic medium.Each fuel element contains thousands of micro-spherical particles.Each particle is dispersed randomly in the space,meanwhile all particles cannot be overlapped.It is infeasible to solve the problem with the explicit particle modeling within the deterministic method.Therefore,a method called the Sanchez-Pomraning method is adopted as following.The particles and matrix are homogenized equivalently with the respective chord length distribution in the particle and matrix.The homogenized medium is addressed with the specific MOC.And the scalar flux of particles and matrix is recovered in a specific way.In this scheme,the particle and matrix chord length is introduced to the neutron transport equation.The MOC method for the stochastic medium fuel is realized,and the corresponding calculation program is termed as ALPHA-DH,which is abbreviated for Advanced Lattice Physics code based on Heterogeneous Architecture-Double Heterogeneity,is written to solve the problem.The accuracy of the program ALPHA-DH is verified by calculating a series of macroscope transport issues.These cases contain various sizes of particles,different packing fractions of particles,multi-type particles co-existing problems,FCM fuel assembly in the PWR and the particle size following a certain distribution caused by the fabrication tolerance.The numerical results show that the ALPHA-DH code can address the stochastic medium fuel issues as a reliable transport solver in a high accuracy.