Bin Zhang

TL;DR: It is the first time to employ the Variational Nodal Method (VNM) as the neutron diffusion solver in PWR core calculation and can eliminate the large error in pin-power reconstruction due to the elimination of the technique in VNM.

TL;DR: In this article, a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) has been developed and applied to investigate the system transient performance during a startup from zero cold power to full power.

TL;DR: In this article, a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model.

TL;DR: The thermal–mechanical model is developed to develop a core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module to analyze the fuel mechanical behavior of fuel rods in different enrichment areas of CAP1400 reactor to provide a basis for judging cladding rupture in severe accidents.

TL;DR: In this paper, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code, which consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer mode.

TL;DR: In this paper, the development and technologies of micro heat pipe cooled reactor are overviewed, and difficulties and challenges need to be overcome in the future, including heat pipe cascading failure, fuel enrichment, structure integrity, machining, monolithic thermal stress, inspection and qualification, etc.

TL;DR: A comprehensive review of sub-channel thermal hydraulic codes used for the analysis of nuclear reactor core is presented in this article, which covers various aspects of experimental, analytical and computational works related to rod bundles carried out in the past and brings out future directions derived from earlier research works.

TL;DR: In this article, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized and experimental data were analyzed and classified, and the inherent mechanism for controversial issues in different experiments was explained.

TL;DR: In this paper, the authors reviewed different N/TH coupling methods, including loose and tight coupling, and classified the current research status based on the coupling methods of neutronics and thermal-hydraulics.

TL;DR: In this paper, the authors proposed a dual-stage Na-TEC, which divides the isothermal expansion into two stages, one at the evaporator temperature (1150 K) and another at an intermediate temperature (650 K-1050 K).