The SM-MSR (small modular molten salt reactor) has a good prospect for development with regards to combining the superiority of the molten salt reactor and modularization technologies, showing the advantages of safety, reliability, low economic cost and flexibility of site selection. However, because its internal structural parts are not easily replaced, and the outer shielding structure is limited, the lifespan of the reactor vessel and its in-reactor shielding design needs to be addressed. In order to find an optimal shielding model with both high fuel efficiency and strong radiation shielding capability, five different design schemes were proposed in this work, which varied in thickness and boron concentration in inner-shielding materials. The neutron/gamma flux and DPA (displacements per atom)/helium production rates were evaluated to obtain an appropriate scheme. Several beneficial results were obtained. Considering the above factors and the actual manufacturing process, 20 cm-thick boron graphite with a 5 wt% Boron-10 concentration combined with a 1 cm-thick Hastelloy barrel was chosen as the in-reactor shielding structure. Outside the reactor, the neutron flux was reduced to 8.33 × 10¹⁰ cm⁻² s⁻¹, and the gamma flux was decreased to 1.13 × 10¹¹ cm⁻² s⁻¹. The vessel/barrel material could maintain a lifespan of more than 10 years, while the burnup depth was 6.25% lower than that of a model without inner-shielding. The conclusions of this research can provide important references for the shielding design and parameter selections of small molten salt reactors in the future.

In molten salt reactors (MSRs), molten salt performs dual essential roles as fuel and coolant. The continuous circulation of the fuel salt in the primary loop inevitably leads to significant neutron activation of loop components, particularly the structural alloys of the heat exchanger (HX) and the coolant salt within the HX. This activation is strongly influenced by delayed neutron fluxes generated by the migration of delayed neutron precursors (DNPs) within the flowing fuel salt. Accurate quantification of the radioactivity of primary HX components is essential for establishing reliable modular replacement strategies, optimizing shutdown maintenance schedules, and ensuring operational safety. To address this requirement, a comprehensive simulation methodology has been developed to model the DNP transport through the primary HX in a small modular molten salt reactor (SM-MSR). It aims to quantitatively evaluate activation levels of HX structural alloys and circulating coolant salt within the HX. Comparative simulations were conducted to contrast scenarios with dynamic DNP migration and static-fuel scenarios excluding it. The results indicate that consideration of DNP migration increases the neutron flux at the top region of the HX by approximately three orders of magnitude compared with the static-fuel scenario. This elevates coolant salt radioactivity by over 50%. Significant increases in irradiation damage parameters (displacements per atom and helium production) are observed in the upper sections of HX structural alloys. These findings highlight the necessity of incorporating DNP migration effects for accurate prediction of primary loop component neutron activation. This provides a reference for future shielding design optimization, irradiation damage assessments, and shutdown dose rate calculations in the SM-MSR.

Accurate specification of spatial heat sources is required for liquid fuel molten salt reactors, given the redistribution of deposited energy among fuel salt, graphite, and metallic structures by gamma ray transport. In this study, a refined power deposition framework based on Monte Carlo particle transport was formulated for a 150 MW thorium molten salt reactor to quantify particle, material, and spatial contributions to the core heat source. The calculation was performed under a static fuel salt approximation, in which flow-induced transport of delayed particle precursors in the circulating fuel salt was not explicitly considered. Particular attention was given to gamma ray generation, transport, and deposition, as well as to the resulting refined power density distribution. The results show that gamma rays contribute 7.64 MW to graphite heating, approximately 3.11 times the neutron contribution. In the alloy sleeve, capture gamma rays account for 63.44% of the deposited power, producing an average power density of 9.34 MW/m³, nearly 18 times that in graphite. This behavior is primarily associated with the strong neutron capture capability of the tungsten bearing alloy and the relatively short gamma ray mean free path in this material. The refined power density distribution further indicates that fuel salt channel power density decreases from about 100.0 MW/m³ near the core center to 57.7 MW/m³ near control rod assemblies, while local graphite power density varies by more than 40%. These findings indicate the need for spatially resolved and material specific heat source descriptions in thermal hydraulic coupling and structural heat load assessment, rather than material averaged treatments or fixed fuel to graphite scaling.