Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

Applicability of RANS models and pressure drop in edge subchannels for 19-pin wire-wrapped fuel bundle channel in CiADS

The accelerator-driven subcritical system has a strong transmutation ability and high inherent safety, and it is internationally recognized as the most promising long-life nuclear waste disposal device. This study involves the construction of a Visual Hydraulic ExperimentaL Platform (VHELP) for the purpose of evaluating the applicability of Reynolds-averaged Navier-Stokes (RANS) models and analyzing the pressure distribution within the fuel bundle channel of China initiative accelerator-driven system (CiADS). Measurements of thirty differential pressures in edge subchannels within a 19-pin wire-wrapped fuel bundle channel were obtained under different conditions using deionized water. The pressure distribution in the fuel bundle channel at Reynolds numbers of 5000, 7500, 10,000, 12,500, and 15,000 was simulated using Fluent. The results show that RANS models obtained accurate results, and the shear stress transport k-ω model provided the most accurate prediction of the pressure distribution. The difference between the results of the Shear stress transport (SST) k-ω model and experimental data was the smallest, and the maximum difference was ±5.57%. Moreover, the error between the experimental data and numerical results of the axial differential pressure was smaller than that of the transverse differential pressure. The pressure periodicity in axial and transverse directions (one pitch) and a relatively three-dimensional pressure measurements were studied. The static pressure fluctuated and decreased periodically as the z-axis coordinate increased. These results can facilitate research on the cross-flow characteristics of liquid metal-cooled fast reactors.

Development of a Single-Phase, Transient, Subchannel Code, within the MOOSE Multi-Physics Computational Framework

Subchannel codes have been widely used for thermal-hydraulics analyses in nuclear reactors. This paper details the development of a novel subchannel code within the Idaho National Laboratory’s (INL) Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a parallel computational framework targeted at the solution of systems of coupled, nonlinear partial differential equations, that often arise in the simulation of nuclear processes. As such, it includes codes/modules able to solve the multiple linear and nonlinear physics that describe a nuclear reactor, under normal operation conditions or accidents. This includes thermal-hydraulics, fuel performance, and neutronics codes, between others. A MOOSE-based subchannel code is a new addition to the fleet of INL-developed codes, based on the MOOSE framework. In this work, we present the derivation of the subchannel equations for a single-phase fluid, we proceed with the description of the algorithm that is used to solve these equations and describe how this algorithm was implemented within MOOSE. We also present how this code can be coupled to the BISON fuel performance code. Next, we verify the friction model and the turbulent mixing model. We calibrate the turbulent modeling parameters for momentum mixing and enthalpy mixing, CT,β. We validate the code using experimental results and last demonstrate the coupling capabilities using a simple example.