We propose a method of burning high-level transuranic waste combining with Th-U fuel cycle in a subcritical reactors driven by external fusion neutron sources. Corresponding one-dimension model is built by means of ONESN_BURN code with new data libraries. Numerical results, including actinide radioactivity, biological hazard potential, and high burn-up rate of high-level transuranic waste are obtained. Comparison with thermal reactor shows that the harder neutron spectrum is the more efficient than the soft one.

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.

With rapid development of computer and super computation, coupling for multi-physics,multi-scale and multi-process has become possible. Some lone process are integrated together. Some approximates from experience will be removed after all of the processes to be considered. This work is based on a virtual reactor. The goal is to improve precision by improvement of modeling and high fidelity computation. At present, study measures are changed from experiment and engineer dependent to theory analysis and numerical simulation. Numerical simulation will become more and more important. In this paper, CASL and NURESAFE are introduced. Then, several challenges, which include key technologies of software, are put forward in development of nuclear energy. Finally, suggestions are given for numerical reactor. It is only for reference and discussion.

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.

Time consuming analysis of grid search in JMCT, a continuous-energy Monte Carlo neutron transport code, is done. Region projecting acceleration is performed on grids of main energy, fission yield neutron and unresolved resonance data. Quality of project grid is shown with arithmetic progression and geometric progression. At last, optimized grid search is test on several k_∞ problems. The models are fuel lattice cell at begin of cycle and middle of cycle. It shows that optimized grid search improves speed of JMCT remarkably in middle of cycle problem.

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.

VENUS-III benchmark was simulated based on 3-D general-purpose Monte Carlo coupled neutron and photon transport code JMCT (J Monte Carlo Transport Code). It includes detailed modeling, criticality calculation and shielding calculation. Firstly, criticality calculation is done, in which k_eff eigenvalue, fluxes in importance regions and energy spectrums are calculated. Almost same results are achieved by JMCT and MCNP. Warp of k_eff is about 0.24% the flux warp is less than 1%. Then, deep penetration shielding is simulated, in which fixed source is used. Tally includes fluxes of detectors in different locations. Calculated results agree well with experimental data. The most warps of JMCT and tests are within target accuracy of ±15%. It satisfies requirement of shielding calculation. It shows that JMCT code suits well for criticality and shielding simulation.

We set up a numerical experiment platform.It consolidates codes used to simulate warhead verification technologies of passive neutron,passive γ ray,active neutron,active high-energy photon and delayed neutron methods.Two functions on simulating time-dependent coincidence and neutron multiplicity counter measurements were added to the platform.They are carried out by DTB code and NMC code.This paper introduces development and validation of the programs,including theory and program flow.Two numerical experiments are designed to validate the program.The platform provides systematic data to support statistical analysis on nuclear warhead verification technologies

A BNCT treating planning system MCDB is developed.Three-dimensional material matrix and tally matrix are designed, which are used to describe voxel models and tally.A fast track technique is used in simulation.Computation time is greatly decreased compared with MCNP code.Same dosimetry results with MCNP are achieved by MCDB,which is faster by 3.1-3.4 times with respect to MCNP.Computation time and accuracy of most voxels reach clinical BNCT requirement in a scale of 10 million particles.

In inertial confinement fusion (ICF) target, collision between fusion neutrons may produce super-high energy neutrons. A relationship between velocity of fusion and ratio of generated super-high energy neutron is derived. An ideal fusion model is simulated by numerical method. It indicates that the ratio of generated super-high energy neutron is increased sharply as burning volume of DT is decreased. With this information, 2D-effect and mixing-effect during compression of target are deduced with escaped super-high energy neutrons. Furthermore, a relationship between velocity of burning T and ratio of generated super-high energy neutron is concluded. It shows that using super-high energy neutron to detect ICF is effective.

An adding weight sample method is developed for neutron transportation in inertial confinement fusion.It aims to increase samples in important zone and decrease samples in non-important zone.Weight of each sample is corrected for keeping calculated results non-bias.Typical model is simulated.It indicates that the method improves effectively neutron samples in fusion zone,while collisions between neutrons increase obviously.Calculation error of super-high energy neutron flux is decreased remarkably.Calculation efficiency of super-high energy neutrons is improved.

Development of Monte Carlo method is over sixty years.It is widely used in nuclear science and other relative fields.It has superpower capability in simulating various complicated geometry.With precision point-wise cross-section,it can simulate various particle transport problems,such as neutron,photon,electron,α particle and proton etc.With rapid development of computers and large scale parallel computation,Monte Carlo method is a first tool for simulating particle transport problems.

Relationship between super-high energy neutrons and collisions is derived.Monte Carlo method is used to decide speed of neutrons after collision.A code applicable to nonlinear neutron transport problem is developed to simulate neutron collision and transportation.Numerical calculation indicates that production rate of super-high energy neutrons is in linear relationship with the square of neutron source intensity.

An approximate linearization method is proposed for nonlinear neutron transport equations. The equation is deduced to a form suitable for Monte Carlo simulations. Numerical results show that it is reasonable. It provides a tool for simulating high energy neutron transportation.

We introduce a simple method for voxel constructing and material ascertainment in BNCT(Boron Neutron Capture Therapy). The voxel model of a Snyder head phantom based on four materials, provides a good approximation. It keeps the conservation mass of skeletoncranium, brain and skin. The 4-material voxel model and 286-material voxel model with mesh sizes of 16 mm, 8 mm, 4 mm, are simulated respectively by the MCNP Monte. Carlo program. The result indicates that the 4-material voxel model is accurate. We also recommend a 5 mm mesh voxel model, which saves simulation time and keeps a good accuracy.

In order to obtain more reaction information of neutron or the other secondary particles by forcing the collision to occur within a small time step, three Monte Carlo Time Forced Collision (TFC) sampling methods called discrete, large cross section and uniform time sampling method are presented, derived and tested. TFC method can be used in more complicated eases. For example, particle transporte through continuously changing materials within a cut time. Numerical results of simple multiplication system show that the calculation efficiency of TFC method is higher than that of the analog method by 2~4 magnitudes.

A parallel algorithm for time-independent Monte Carlo transport is successful since particles are independent and they are distributed to multiple processors.However,for time-dependent Monte Carlo transport problems, the parallel efficiency reduces and the parallel scale is limited due to the communication of scattering source attribute and meshes in each time-step.We propose two algorithms in them adaptive processor assignment and optimized processor choice are obtained.With a Monte Carlo stratified sampling technique for scattering source treatment the communication cost is reduced greatly.The parallel expandability is improved.A large speedup over the basic algorithm is obtained.

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 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.

It is a valid method by using the Monte Carlo method to calculate three-dimensional porosity well-logging problems. This paper gives out the calculated results of the benchmark neutron well-logging problems with porosity 1% and 20% from the MORSE-CG code which the restart capability have been added into it, and thd comparison of it with MCNP and McDNL results and computing time. Results indicate that the three cokes give the same results within calculated standard deviations.

This paper introduces the combinatorial geometry package CGT with the capability of the torus geometric treatment. It is get by developing the combinatorial geometry package CG. The CGT package can be transplanted to those codes which the CG package is being used and makes them also with the capability. The MORSE-CGT code can be used to solve the neutron, gamma-ray or coupled neutron-gamma-ray transport problems and time dependence for both shielding and criticality problems in torus system or system which is produced by arbitrary finite combining torus with torus or other bodies in CG package and it can also be used to design the blanket and compute shielding for TOKAMAK Fusion-Fission Hybrid Reactor.