An in-core fuel management code package HEX-ICFM is developed for Xi'an Pulsed Reactor.The orthogonal design model and the loading pattern optimization program HEX-ORTH are studied.The cell calculation is performed using WIMS-D/4.A two dimensional hexagonal geometry multigroup nodal theory code SIXTUS-2 is used for core diffusion calculation.The core physical parameters are calculated using HEX-ICFM for the first cycle of Xi'an Pulsed Reactor.With the fuel of end of cycle 3 and 30 fresh fuel rods the optimum loading patterns are calculated for the objective function max(K^BOC_eff)using HEX-ORTH.

A 3-D multigroup P₃ approximation Monte Carlo code MCMG-BURN is developed by coupling the neutron transport and burnup.MCMG-BURN code is based on the continuous-energy cross-section Monte Carlo code MCNP and the lattice homogeneous code WIMS.It uses the multigroup cross-section libraries to simulate the critical test reactors and HFETR(High Flux Engineering Test Reactor).The agreement results with the MCNP results and experiments are achieved.The MCMG-BURN code is at almost the same precision with the MCNP code while requiring considerably less computing time.

The Statistical Estimator (SE) and the Expected Value Technique (EVT), which belong to the range of Monte Carlo method, have been introduced and are used to simulate the coupled neutron and gamma transportation.The total γ energy spectrum,the time spectrum and the characteristic γ spectrum of each element of the detected objects can be obtained respectively by using the two techniques.The validation of the new method is proved by the comparison with the results of the coupled neutron and gamma Monte Carlo code MCNP.

When the fission reaction happens in neutron transport calculation,the energy of the fission neutron can be usually determined by the fission spectrum.Since the coefficient of the spectrum type depends on the incident neutron energy,the multigroup neutron fission spectrum should strictly be a matrix form.In general transport calculation,the ²³⁵U vector fission spectrum is usually chosen as the standard fission spectrum.In order to make clear what kind of effect will be produced after this treatment,the Monte Carlo method and the multigroup P₃ Monte Carlo neutron transport code MCMG are used to analyze the difference between two types of fission spectrums.By comparison the results with MCNP code,it is certain that the difference exists,but it does not have any effect on the correction of the calculated results.At the same time,the calculation results of the different neutron cross section libraries have been compared.

The sampling method by coupling criticality source direction biasing and the exponential transform is established for duct shielding calculation.In sample problem studying,the optimal source direction biasing parameter p₁、exponential transform parameter p₂ are supposed and the variance of duct calculating results by using this method is shown to reduce efficiently.Correction calculation for the duct of Nuclear Power Institute of China (NPIC) pulsed reactor shows the results are more close to the measured ones.

The parallel design of Monte Carlo code involves computational method and module designs,which is crucial to the parallel efficiency.The coupled of neutron and photon transport Monte Carlo code MCNP has been realized the parallelization in PVM and MPI by modifying the serial code.Due to the form having module being optimized, the parallel efficiency is good where the efficiency of MPI code is stronger than that of PVM code and the speedup of MPI code is higher than that of PVM in most cases.The calculated results of parallel code are reasonable.Both the speedups of PVM code and MPI code have been the linear increasing with the processors.The parallel efficiencies are up to 99% in 16-processors,97% in 32-processors and 89% in 64-processors respectively.

It outlines MCMGP3-A 3-D multigroup P₃ Monte Carlo neutron transport code, which is designed for computation of reactor critical safety. It is developed from the coupled neutron and photon transport Monte Carlo code MCNP with continual point cross sections and the continual point cross section module of MCNP code has been replaced by multigroup cross section module. MCMGP3 code is with the capablilities of MCNP geometry treatment, counting, deviation reducing techniques and plot. Either macroscopic or microscopic cross sections can be used.The neutron scattering angular distribution adopts P₃ approximation and generalized Gaussian quadrature scheme. The sample problem results prove the computational correctness of the code. MCMGP3 code can treat steady or unsteady porblems with exterior source and critical problems. It suits various cross section formats and the energy groups can be expanded easily from one group to multigroup. In addition, the MCMGP3 code has successfully realized the coupled with the reactor lattic code WIMS. It can be used to simulate the full reactor problems.