Session: 07-13: SMR and Advanced Reactors - II

Paper Number: 136580

136580 - Study on Thermal-Hydraulic Characteristics of Horizontal Lead-Based Reactor Assembly Under Complex Motion Condition 

Abstract: 

Lead-based Fast Reactor (LFR) has high power density, strong inherent safety, and excellent maneuverability, which can meet the requirements of high efficiency, stability, safety, and long endurance of deep-sea equipment. During the operation of the Marine Nuclear Power Plant (MNPP), the thermal-hydraulic and safety characteristics of the horizontal reactor core are different from those of the land-based reactor due to the complex ocean environment. Therefore, the corresponding analysis codes need to be urgently developed.

According to the special structure and operational characteristics of horizontal LFR under marine conditions, the Subchannel Analysis Codes Of Safety (SACOS-PB) was developed in this paper. With SACOS-PB code, the thermal-hydraulic characteristics of a single assembly in horizontal LFR were analyzed, mainly focusing on the difference between horizontal assembly and vertical assembly in flow and temperature distribution. The best important input parameters and models were obtained for the whole core simulation, and the influence of typical ocean conditions on thermal parameters was obtained. The flow direction of the coolant is perpendicular to gravity in horizontal assembly. The steady-state analysis shows that compared with the centrosymmetric distribution of coolant flow and temperature in the vertical assembly, the flow in the horizontal assembly increases along the gravity direction, and the temperature rises first and then decreases. The maximum coolant temperature in the section is relatively higher in horizontal assembly. In the steady-state analysis, the parameter and model sensitivity of cross-flow resistance factor, gap heat conduction factor, turbulent mixing factor, heat transfer coefficient model, and friction coefficient model were analyzed, and the best parameter value or model is recommended for transient calculation and assessment. Three typical transient conditions were considered in the marine conditions, including inclined, heaving, and rolling conditions. The axis of inclined and rolling motion could be the x- or y-axis. In the inclined conditions, the gravity component changes, which affects the crossflows between adjacent channels. The coolant flow and temperature distribution are mainly influenced by the variable crossflows. The maximum coolant temperature at the assembly outlet increases with heeling angle and decreases with the trimming angle. The thermal-hydraulic parameters of the coolant fluctuate periodically under heaving and rolling. The fluctuation period of flow and temperature is consistent with that of marine conditions. In the above transient conditions, the maximum coolant temperature at the assembly outlet increases. The crossflow perpendicular to gravity is more sensitive to ocean motion. The coolant flow and temperature distribution are significantly affected by the inclined and rolling with the x-axis, which is greater than the motion with the y-axis.

With SACOS-PB code, the thermal-hydraulic evaluation of the whole core in horizontal LFR was conducted, mainly focusing on the flow and temperature distribution of the hottest assembly. The influence of coupled ocean conditions on thermal parameters was obtained. In contrast to the single assembly analysis, the flow and temperature difference between the upper and lower of the hottest assembly is significantly reduced due to the crossflow mixing with adjacent assemblies. Three ocean conditions were considered, including heaving, rolling, and coupled motion. Single ocean motion in the coupled transient condition will affect the thermal parameters of the core. The larger the marine motion amplitude is, the stronger the parameter fluctuation is. The period of ocean condition not only changes the period of the flow and temperature but also affects the fluctuation peak and valley values.

Presenting Author: Minyang Gui Harbin Engineering University

Presenting Author Biography: GUI Minyang received his Ph.D. in Nuclear Science and Technology from Xi’an Jiaotong University, in 2022. Upon graduating, he took a position as a Lecturer in the College of Nuclear Science and Technology at the Harbin Engineering University. His research interests include multi-scale numerical simulation of nuclear reactor (including system analysis, pin-by-pin sub-channel analysis and three-dimensional CFD analysis), development of critical heat flux mechanistic model, and thermal hydraulic characteristics analysis under ocean conditions. He has presided over or participated in many research projects such as the National Natural Science Foundation, National Major science and Technology project and some horizontal projects from public institution. He has published more than 20 papers. At present, he is the reviewer of well-known journals such as Progress in Nuclear Energy, Front. in Nucl.Eng.-Fission and Reactor Desg.

Authors:

Minyang Gui Harbin Engineering University
Di Wu Harbin Engineering University
Hangyuan Zhang Harbin Engineering University
Senmiao Zhang Harbin Engineering University
Hui Cheng Harbin Engineering University