Search

The helically metallic fuel is promising to improve the power density and safety margin of reactors core by its advantages of high thermal conductivity, large heat transfer area-to-volume ratio and continual inter-channel mixing. However, helical geometry and metallic U-Zr fuel will introduce challenges in the analysis of neutron physics, thermohydraulics and mechanics performance. Thus, in the current work, the analysis methods were developed for the characteristics of helically metallic fuel in neutron physics, thermohydraulics, mechanics and multi-physics coupling. For neutron physics, a 3D continuous-energy Monte Carlo method was employed to address the complex geometries and the complex intermediate neutron energy spectrum and generate few-group cross-sections. In core calculations, the 3D method of characteristics (MOC) with high geometric flexibility, accuracy and convergence was utilized. This method has been verified in 3D calculations of helically metallic fuel core. The results demonstrate that both cross-section generation and core calculation have achieved high accuracy and effectively improved computational efficiency. For thermohydraulics, the experimental technique including visual measurement and wire mesh sensor were used to measure the mixing characteristics of single and two-phase flow in helically metallic fuel rod bundle. CFD and subchannel analysis codes were developed to predict the boiling and critical heat flux. The inter-channel mixing was dominated by the flow sweeping mixing, while the turbulent mixing was insignificant. The vapor phase crowded at the elbow of the rods, where the boiling crisis was triggered. For mechanics, the molecular dynamics method was employed to predict the fundamental thermal-mechanical properties. The correlations were proposed for the thermal conductivity and elasticity modulus of U-Zr alloy under irradiation conditions with pores. The finite element tool Abaqus was used to analysis the thermal-mechanical performance of helically metallic fuel rod bundle. The maximum stress was found at the blade tips where adjacent rods contact with each other. The maximum stress varied from 326.7 MPa under fresh condition to about 313.3 MPa at 14.1% FEMA because of creep, which was far lower than the failure strength of the rod. The structure integrity can be ensured during the full life cycle. Based on the individual analysis method, the multi-physics coupling framework combining the neutron physics, thermohydraulics and mechanics was developed for the design and performance evaluation of the fuel assembly and reactor core with helically metallic fuel rods. Based on this multi-physics analysis tool, a novel boron-free small modular pressurized water reactor NETH-HCF175M design using helical-cruciform metal fuel was proposed. The core can achieve a cycle length of about 1 360 EFPD as the reactivity of the core is completely controlled by burnable absorbers and control rods.

BackgroundPassive residual heat-removal system (PRHRS) based on natural circulation has been widely used in small reactors connected on secondary side loop.PurposeThis study aims to develop and analyze a passive residual heat-removal system on the secondary side of a small reactor to improve reactor safety.MethodsBased on the completed heat-transfer experiment of the intermediate circuit of a small integrated nuclear power plant, Reactor Excursion and Leak Analysis Program (RELAP) was employed to determine and analyze the natural circulation characteristics of the intermediate loop were determined and analyzed.ResultsThe heat transfer rate results of the program are in good agreement with the experimental data, and the natural cycle characteristics of PRHRS can be characterized. The pressure of the system loop is determined by the average temperature of the primary side of the steam generator (SG), and the difference in the inlet temperature of the SG primary side, mass flow rate, and height of the cooling and heating sources significantly influence the heat-transfer performance of the SG system. The heat-transfer performance of the waste-heat-removal system is more sensitive to the resistance of the system loop when the temperature of the primary-side inlet of the SG is higher.ConclusionsThese results provide a valuable application for further investigations of passive systems in small reactors.

Subchannel analysis is one of the most essential methods in the thermal design of nuclear reactor cores and transient safety analysis. Spacers are arranged to maintain proper geometrical configurations of rod bundles and contribute to crossflow mixing in rod bundles. In this work, a general flow sweeping mixing model is introduced for subchannel analysis, and the model is applicable for rod bundles with arbitrary spacer types. Applications of this model to the mixing vane spacer grid rod bundle, the helical fuel rod bundle, and the wire-wrapped rod bundle are discussed. The degenerate crossflow rates downstream of the mixing vane spacer are accurately predicted, demonstrating the superiority of the GFSM model. When applied to the helical fuel rod bundle, the refined subchannel division approach should be utilized together. The GFSM model can predict twist angle-dependent crossflow rates, showing better agreement with CFD results compared to the existing flow sweeping model. Furthermore, the crossflow rates predicted by the GFSM model in the wire-wrapped rod bundle exhibit higher prediction accuracy. This work provides a new approach to develop the flow sweeping mixing model and unifies the model forms for different rod bundle types. [ABSTRACT FROM AUTHOR]