Heat pipe cooled reactors rely on heat pipes for passive heat transfer and exhibit high reliability and compactness. Therefore, they are considered candidate nuclear reactor systems for future deep space exploration missions. To enable a deeper investigation of heat pipe reactor systems, particularly the transient response characteristics of the core, a transient coupled analysis framework is developed based on the multi-physics coupling code MOOSE. This framework includes the core heat transfer module, point kinetics module, heat pipe module, and Stirling engine module. A novel strategy that allows two distinct heat pipe models to be simultaneously invoked within a single simulation in MOOSE is developed. All modules are developed within the MOOSE framework and do not rely on any external programs. The heat pipe module is validated using experimental data from heat pipe startup and operation tests within the maximum relative error of only 0.45%. The entire coupled framework is validated against the KRUSTY operational experiments and is compared with other multi-physics models, demonstrating higher accuracy within the maximum relative error of only 13.7% in core load variation conditions. Meanwhile, transient coupled analyses of the KRUSTY reactor are performed to evaluate its safety performance under accident conditions. In the hypothetical positive reactivity step insertion accident and heat pipe failure accidents, the KRUSTY core exhibits excellent safety performance. And the mechanism of heat pipe power redistribution following heat pipe failure is examined in detail.

A design for a small chloride salt fast reactor (sm-MCFR) is presented through the integration of molten salt reactor and small reactor technologies, targeting efficient transmutation of transuranic (TRU) elements in spent nuclear fuel and rapid reactor deployment. The feasibility exploration and research on the design boundaries of sm-MCFR will be conducted in this article. The core adopts a dual-fluid configuration, in which the fuel salt and coolant circulate independently. Chloride salt is selected as the fuel carrier due to its high solubility for heavy metal nuclides and the low neutron absorption cross-section of chlorine, which help to form a hard fast-neutron spectrum and thereby enhance transmutation efficiency. The cooling system employs a direct supercritical carbon dioxide (s-CO₂) cycle, simplifying the overall layout. For the neutronics design, simulations were carried out using the TMCBurnup (TRITON MODEC Coupled Burnup Code). By adjusting the core geometry, fuel salt composition, and reprocessing strategy, the sm-MCFR achieves a hard fast-neutron spectrum but also demonstrates good potential for fuel utilization. In terms of thermal–hydraulic design, the heat exchange effect of the reactor core can be improved by adjusting the proportion of the coolant and the flow direction. The sm-MCFR is expected to become a promising candidate for advanced small reactors that have potential applications in nuclear waste transmutation and distributed energy generation.

Conventional dispersed-particle fuel systems, such as TRISO fuel, exhibit typical double heterogeneity. The Reactivity-equivalent Physical Transformation (RPT) method can convert a double-heterogeneity system into a single-heterogeneity system by simplified geometric modelling. In a Pebble-Bed Fluoride-Salt-Cooled Reactor (PB-FHR), the random packing pebbles and the fluoride-salt coolant further impacted the spatial self-shielding effect, fuel loading and moderating ratio, causing the variation in the RPT radius. In this study, a neutronics model of the TMSR-SF1 pebble lattice was established with OpenMC. The VHM, RPT and random TRISO models were compared and validated from the perspective of neutronics. The effects of pebble lattice structure, pebble packing fraction, FLiBe coolant temperature, TRISO loading fraction, various burnable neutron poisons and boundary albedo on RPT radius were systematically evaluated. Results indicated that RPT radius is unaffected by the pebble packing structure, while it was positively correlated with pebble packing fraction or coolant temperature and was negatively correlated with the moderating ratio. Poison with strongly neutron absorption may reduce RPT radius, while poison with weakly neutron absorption has little effect on RPT radius. Reducing neutron leakage will lead to an increase in a RPT radius. This work would provide an effective view for the improvement of RPT method in PB-FHR, and as the basis for the subsequent full core analysis.