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A transportable fluoride-salt-cooled high-temperature reactor (TFHR) design with 20-MW(thermal) rated power and 18-month fuel cycle is proposed for off-grid applications. One of the design goals of the compact reactor core is potential transport by truck, rail, or air. Full-core thermal-hydraulic analyses and improvements using three-dimensional computational fluid dynamics (CFD) were performed previously to demonstrate the feasibility of a TFHR design at a nominal power of 20 MW(thermal). In this paper, the best-estimate system code Reactor Excursion Leak Analysis Program (RELAP5-3D) is adopted to study the transient behavior of this TFHR design and the safety characteristics of the primary loop system during accident conditions. The modeling results of the steady state were verified using CFD results with consideration of radial heat conduction between heat transfer unit cells. Four most challenging accidents of anticipated transient without scram were analyzed, as well as parametric studies of some key factors. These accidents include unprotected reactivity insertion accident (URIA), unprotected loss of heat sink (ULOHS), unprotected loss of flow (ULOF), and a combination accident of ULOF and ULOHS. The results indicate that transient temperature limits are not exceeded during the most severe accidents. They indicate satisfactory transient performance of the TFHR design. The transient temperature limit of structure material Hastelloy N, based on embrittlement phenomena, poses the most limiting constraint due to the small temperature margin of about 20 K in the accident combination of ULOF and ULOHS. Overall, TFHR is a sound reactor design from a thermal-hydraulic viewpoint. [ABSTRACT FROM PUBLISHER]

Reliable systematic uncertainty analysis is of great significance for reactor safety and licensing. In this paper, efforts have been made to quantify uncertainties of computational results. The uncertainty quantification and sensitivity analysis for a 20 MWth prismatic Transportable Fluoride-salt-cooled High temperature Reactor (TFHR) proposed by MIT is performed with the DAKOTA code coupled to the best-estimate nuclear system software RELAP5-3D. Eight input parameters are selected with specified variations and probability density functions (PDFs), a total number of 400 steady-state and Unprotected Loss of Flow (ULOF) transient calculations are then conducted in order to statistically obtain the 5% and 95% peaking fuel temperature tolerance intervals with confidence level of 95%. The Latin hypercube sampling (LHS) model in DAKOTA is verified against analytical results and the TFHR RELAP5 modeling is benchmarked based on CFD calculations. It is found that, for the steady-state, the uncertainties of the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. 0.5% of calculated fuel temperatures exceeded the steady-state fuel temperature limit of 1573 K. Under ULOF accident condition, no temperature points exceeded the transient fuel failure limit 1873 K, but there is a possibility of 0.195 leading to the damage of structural material. Three primary contributors to the uncertainty in peak fuel and outlet coolant temperature are the reactor power, helium gap, and heat transfer coefficient. The findings indicate that there are needs for further reactor design optimization from thermal-hydraulics prospective. [ABSTRACT FROM AUTHOR]