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A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies. [ABSTRACT FROM AUTHOR]

A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies.

The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan’s specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident. The design concepts of a reactor vessel (RV) and its internal structures have been investigated whether they could withstand severe seismic conditions in Japan and thermal loads. The design adopted enhanced RV support structure, enhanced conical-shaped core support structure, a thickened knuckle part of the RV, and a flat plenum separator with ribs. A three-dimensional steady-state thermal hydraulic analysis of the RV revealed that the temperature difference of the upper and lower surfaces of the flat plenum separator could be effectively reduced by installing layers of thermal insulation plates. The authors have also conducted a transient analysis of loss of flow type anticipated transients without scram events to evaluate the feasibility of a self-actuated shutdown system. Moreover, the configuration of the decay heat removal system has been investigated considering sufficient utilization of natural circulation capability of sodium coolant, heat removal capacity of each cooling system, conformance with design requirements, and recommendations of SDC and SDG such as diversity and redundancy of components.