[en] This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

[en] The theoretical model and its application of CELL code are introduced. The code solves the integral transport equation in concentric circular geometry by collision probability method to obtain neutron spectrum and few group parameters. For a fuel rod cell and a burnable poison rod cell, the deviations of K obtained by WINS-D/4 and CELL are 0.024% and 0.23%, respectively. For the five zero power experiment cores, concerned with water reflector or beryllium reflector, the deviations of core K_eff measured value and calculated value with CELL and CITATION codes are less than 0.5%. For the first core of HFETR, the control rod positions calculated values of cold zero power and hot zero power are agreed with the measured values very well

[en] Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

[en] The TPLIB-94 library was produced for LWR fuel assembly calculation program TPFAP, based on the JENDL-3.1 evaluated nuclear data library. The calculation results and the measurement values of the 5 thermal reactor benchmark problems and a set of PWR critical experiments are presented. The comparison of these results show that the deviations of the core k_eff values calculated using TPLIB-94 and the experiment values are less than 0.4%. they are obviously less than that of the values using old library

[en] The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

[en] A code RCS-I (Reactor Core Simulator) for reactor core 3D physics-thermal hydraulics couple calculation is introduced. The neutronics model in the code is an advanced coarse nodal Green's function method, and the thermal hydraulics model is a subchannel analysis method. By using several feedbacks, the code can more really describe the burnup feature of a core. It has criticality, burnup, poisoning, load trace and refueling calculating functions, and can be used in the power reactor and research reactor design

[en] The simplified spherical harmonics (SP_N) method is utilized to discretize the angular variables of neutron transport time-space kinetics equation. The finite element method, fully implicit scheme and direct analytical time integration method are used to deal with the spatial, time variables and delayed neutron precursor equation, respectively. According to the model, a computer program is developed to solve multi-dimensional time-space kinetics equation in unstructured-meshes. The numerical results show that this method can be used to perform the kinetics calculation in complex environment. (authors)

[en] The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

[en] The MCATHAS system of coupled neutronics/thermal-hydraulics in the supercritical water reactor is described, which considers the interaction between the obvious axial evolution of material temperature and density and the power distribution. This code is coupled externally. The MCNP code with the library of continuous cross section is used for neutronics analysis. The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN code for burn-up analysis. The calculation results for the assembly of HPLWR show that the results from this code is reliable. (authors)

[en] Based on the conformal mapping, Green function method was applied in hexagonal geometry. Conformal mapping was used to map a hexagonal node to a rectangular node before transverse integration. Then, the transverse integration equations were resolved using Green function method with the second boundary condition. A three dimensional multi-energy-groups static program NACK was programmed based on those theories. The code was verified by VVER-1000-type core without the reflector, VVER 440-type three-dimensional two-energy-groups core and two-dimensional core with discontinuity factors. The eigenvalue error is less than 50 pcm, and the maximum relative error of the node average power is less than 2%. The accuracy of NACK is as good as that of other advanced node method codes. (authors)