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.

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

An improved sliced sensitivity reconstruction algorithm is proposed for three-dimensional(3D) magnetic resonance electrical impedance tomography(MREIT).Using single component of magnetic flux density measured by a MR scanner,all slices are reconstructed respectively with a modification for 3D conductivity images.Simulation of a cubic model shows that the proposed algorithm improves interlayer crosstalk phenomenon with a higher image resolution than sliced reconstruction without modification and entire reconstruction algorithm.Computing time is significantly reduced.Simulation on a human leg model shows feasibility for complex imaging objects.Phantom results verify reconstruction algorithm in real measurements.The improved sliced sensitivity reconstruction algorithm reduces requirements for computer capability and reconstruction time of 3D MREIT with high imaging precision and efficiency.

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.

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.

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.