Coupled Neutronics/Thermal-hydraulics Simulation And Safety Analysis Of Supercritical Light Water Reactor

For supercritical light water-cooled reactor, it is characterized as lower flow rate design, narrower fuel rod gaps and larger temperature increase from inlet to outlet of the core than that of traditional light water reactors, which bring new problems for its safety analysis especially with the influence of neutron-thermal coupling and coolant supply. According to the coupling and safety codes, the research on coupling between thermal-hydraulic and neutronic is carried out, and the safety analysis is put forward focusing on the transients or events of supercritical water-cooled reactor associated with coolant flow rate changing. For the coupling characteristics analysis, supercritical narrow effect was firstly brought forward. And for the first time in China, the coupling method was introduced into safety characteristics analysis of supercritical water-cooled reactor.Firstly, the research on steady coupling characteristics analysis as well as supercritical narrow effects was given.(ⅰ) By comparing the coupling results with that results given by uncoupling method with cosine approximated, it shows that neutronic and thermal-hydraulic coupling will lead to the axial power peak factor shifting along axial direction for both the internal and external components. It makes part of the cladding temperature rising but the maximum cladding temperature decreasing.(ⅱ) For given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor in different gap sizes is given by changing the fuel rod pitch only, and the characteristics of narrow channels effect with enhancing heat transfer are proposed. As the gap size decreases, maximum cladding temperature decreases while the convective heat transfer coefficient between fuel rod and coolant will be risen. Based on thermal analysis in different coolant flow rate, the reasonable value range of gaps between fuel rods is given, in which the maximum cladding temperature safety limits and installation technology are comprehensively considered.Secondly, studies on the transient coupling characteristics analysis as well as the thermal characteristics during start up are put forward.(ⅰ) Transient physical characteristics analysis is given by changing coolant temperature as well as moderator temperature. It shows that, as the coolant temperature increases, the core power decreases with the coupling influence. While the moderator temperature increases, it gives the same results but with smaller decreased degree.(ⅱ) Transient thermal characteristics analysis is given by changing core power. It shows that the coolant temperature increases while maximum cladding temperature decreases but with the inhibition by coupling influence.(iii) According to the given slidding pressure mode of SCWR, the transient characteristics for power increasing is detailed analyzed. Considering with radial heterogeneity of power distribution, thermal characteristics for different assemblies are also put forward. Then on the principle of startup safety and stability, optimization design of slipping startup for supercritical pressure light water reactor is given by adjusting flow rate distribution in different fuel assemblies or changing power setting during startup.Finally, safety characteristics analysis is given.(i) The comparative analysis of transient safety characteristics with main parameters disturbance is made by taking the feed-water flow rate decreasing, temperature decreasing and pressure increasing for example. It shows that the feed-water flow rate decreasing and temperature decreasing gives serious variation of system characteristics (ii) Considering the importance of coolant supply for supercritical water-cooled reactor, transient analysis of single coolant pump failure rate is given by coupled neutronics and thermal hydraulics calculation method, as well as that of5%partial loss of coolant flow rate. On this basis, sensitivity analysis is performed. The results shows that maximum cladding temperature calculated by coupling method is always lower than that calculated by uncoupling method (iii) By the example of Loss of Feed-water Heating Transient and Partial Loss of Reactor Coolant Flow Transient, multi-channel safety analysis is given by introducing the radial power distribution factors and flow distribution calculation. In which, the axial power distribution factors is obtained by curve fitting based on the coupled calculation results. During the above two transient process, there is a peak value of maximum cladding temperature exists in each fuel assembly, in which the maximum cladding temperature peak of fuel assembly with the highest power factor is much higher than that of other fuel assembly. But this situation still meets the safety standards.