Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

Abstract

A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

Status and Assessment Method of Nuclear Safety Analysis Software in China

“…Universities in China have also developed many software applications. For example, the research group NuThel from Xi'an Jiaotong University has developed a series of thermalhydraulic analysis software programs (Jie et al, 2013); (Shiying et al, 2018); (Tian et al, 2007); (Xiao et al, 2014), and the NECP team, also from Xi'an Jiaotong University, has developed a complete set of physical safety analysis software programs (Xu et al, 2017); (Hongchun et al, 2019); however, these pieces of software have not been developed for safety analysis of nuclear power plants, are short of complete quality assurance, and have not been certified and validated sufficiently. Therefore, at this stage, NNSA will not evaluate these programs.…”

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

“…The thermal-hydraulic variables of subchannel analyses of plate type fuel reactors found in the literature are coolant, cladding, and fuel axial temperature distributions, fuel assembly flow rate distribution, pressure drops across the reactor core, mean coolant velocity in subchannels, and departure from nucleate boiling results. Below we describe the IEA-R1 research reactor and the CARR multipurpose reactor and their reported subchannel analysis data [7,22,26,28,34] and the approach taken in this work to carry out the COTENP code validation. Figure 6 shows the top view of configuration 247 and more information about the IEA-R1 reactor can be found in Appendix A.…”

“…Position (0,0) multipurpose reactor, a 56 MW reactor with slightly pressurized light water as the primary coolant and moderator and heavy water as reflector [28,34]. Figure 7 displays a top view of the CARR reactor and detailed information about its geometry and operational conditions is found in Appendix A. Thermal-hydraulic analyses for the CARR reactor under steady and transient conditions were carried out by Lei and Zhang [7].…”