[en] The LOFA and SBO severe accidents are simulated with the ASTCE code separately. The reaction mechanism between metals (Zr, Fe, B₄C) and steam is described in LOFA. The mass and rate and the starting time of hydrogen production are compared. It shows that the hydrogen mainly come from the Zr oxidation, and the hydrogen from Fe oxidation is only 10% of all. The hydrogen production during the accidents of LOFA and SBO are compared also. The mass, the maximum rate and the starting time of hydrogen generation of the same reactor are quite different in different severe accident sequence. So it is necessary to make the analyses of the hydrogen source term during the severe accident and take some effective measures. (authors)

[en] The LOFA and SBO severe accidents are simulated with the ASTCE code separately. The reaction mechanism between metals (Zr, Fe, B₄C) and steam is described in LOFA. The mass, rate and starting time of hydrogen production are compared. The results show that the hydrogen mainly came from the Zr oxidation, and the hydrogen from Fe oxidation is only 10% of the total hydrogen. The hydrogen production during LOFA and SBO accidents are also compared. The mass, maximum rate and starting time of hydrogen generation of the same reactor are quite different in different severe accident sequence. So it is necessary to analyze the hydrogen source term during severe accident and take some effective measures. (authors)

[en] This paper introduces three different real-time simulation models of reactor core neutron kinetics: point kinetics, three-dimensional adiabatic method and improved quasistatic method. Comparison has been made by solving one-dimensional benchmark problem issued by American Argonne National Laboratory and calculating the rod-eject transient of Qinshan 600 MW Nuclear Power Plant. Considering both the computational accuracy and time, the improved quasistatic method is the ideal model for real-time simulation of reactor core kinetics with contemporary powerful computer

[en] During the past several years, the Chinese nuclear industry has seen rapid development and more applications of nuclear simulation technology. This article summarizes the real time simulation technology used for nuclear engineering in Research Institute of Nuclear Research Institute (RINPO). RINPO has developed its latest simulation technology in two key aspects, one is the general Linux-based simulation platform RINSIMTM, which consists of real time development and operating environment, the instructor station, the operator stations and all graphical object-oriented modeling tools. The other is the modeling technology, including five-equation thermalhydraulic code, real time RELAP5 simulation code, 3D reactor physics code, and DCS Emulation solution. The advanced simulation technology makes the simulator development less labor intensive, better maintainability, and high fidelity assured. (authors)

[en] The MAAP5.03 and 5.04 codes is applied to model the CP300 NPP. For the same system model, taking the sequence of large break accidents in the cold leg of primary system for an example in this paper, the differences between MAAP5.03 and 5.04 on core behaviors, hydrogen source terms, steam generator and containment responses are studied. The results showed that the water seal model that is more in line with actual physical is adopted in MAAP5.04 code, it reacts on the heat sink effect on SG secondary side, which makes the differences of steam generator and containment responses; In the meanwhile, because of the heat transfer of natural circulation in primary side, there are some differences between the MAAP5.03 and 5.04 on core behaviors, hydrogen source terms. The relevant data of which can provide important reference for the use and evaluation of MAAP5 code. (authors)

[en] [Background] Based on the latest release of benchmark for evaluation and validation of reactor simulations (BEAVRS), verification of three-dimensional core simulation is one of important issues. [Purpose] This study aims to verify simulation modeling of three-dimensional core in coordinate with BEAVRS version 2.0 and establish corresponding burnup analysis model. [Methods] First of all, 2 D transport theory with multi-group was used to calculate the fuel burnup, and few-group constant of discrete state points for calculation of nuclear cross sections fitting model. Then, the advanced nodal method was used to solve the neutron diffusion equation, and establish the 3 D core nodal code SimOR simulation model. Finally, various operating conditions were selected to carry out component homogenization calculation and core critical calculation. [Results] The calculation results are in good agreement with the measured values of BEAVRS benchmark and the reference results of nTRACER code, which verifies the correctness of the simulation model and the accuracy of the program calculation. [Conclusion] This study provides data basis and scheme reference for the calculation of fuel management components and core diffusion-burnup calculation of PWR. (authors)

[en] In order to optimize and improve the debugging process and method of the distributed control system (DCS) of the nuclear power plant, a non-safety level DCS configuration verification method of the nuclear power plant based on the simulation technology is studied, and its implementation process and key problems are described. The application results of a HPR1000 nuclear power project show that this method does not depend on the actual DCS equipment, and can achieve closed-loop control and integrated test, which is beneficial to shorten the on-site debugging time of DCS and ensure the overall debugging progress of the unit. The non-safety level DCS configuration verification method of THE nuclear power plant based on the simulation technology has certain market promotion value. (authors)