Conference Agenda

Helical Cruciform Fuel (HCF) has a cruciform shape with helically twisted surface, which provides about 35% larger heat transfer area compared to standard cylindrical fuel. From a thermal-hydraulic perspective, this geometry results in a lower wall average heat flux leading to a power uprate potential. This study investigates two key thermal hydraulic-related phenomena for the HCF rods using Computational Fluid Dynamics (CFD) simulation: Departure from Nucleate Boiling (DNB) and fluid-elastic instability. The NuScale-like Small Modular Reactor (SMR) with a low mass low rate is considered as a reference plant design.

First, a numerical boiling test is conducted for a hot fuel pin using a Eulerian-based two-fluid approach, using the CASL boiling models to estimate the Critical Heat Flux (CHF) and Minimum Departure from Nucleate Boiling Ratio (MDNBR). Additionally, post-CHF cladding surface temperature is estimated to provide boundary conditions for a future fuel performance analyses. The performance of HCF is compared with that of standard cylindrical fuel under the same conditions to assess its relative advantages.

Furthermore, a Fluid-Structure Interaction (FSI) simulation is performed to estimate the Fluid-elastic Instability Margin (FIM) in the HCF geometry through vibration analysis. An unsteady simulation is carried out for a 2x2 lattice using the STRUCT-𝜀 turbulence model, which can capture vibrations without the need for LES-level mesh refinement. By using the vibration frequency and damping ratio—computed using the displacement data from the simulation —the FIM and fretting wear rate are estimated.

A global research effort is underway to develop Small Modular Reactors (SMRs) with diverse applications beyond just providing baseload power, while simultaneously enhancing safety. One design of interest is the NuScale concept reactor design, which utilizes natural circulation driven by temperature differences in the primary system, allowing for operation without pumps. This design has attention in worldwide for its high safety profile. However, a significant limitation of the natural circulation reactor concept is the lack of extensive operational experience. To overcome limitations, it is crucial to construct and operating scaled experimental facilities that can simulate natural circulation. In this study, the URI-SMR (UNIST Reactor Innovation-SMR), a scaled-down experimental facility based on the NuScale concept design, was employed to evaluate the natural circulation performance. The URI-SMR is well-suited for natural circulation study because its primary system is constructed of acrylic, which enables simultaneous performance evaluation and visual observation of the natural circulation flow. Through the URI-SMR, steady state experiments at various power levels were conducted, and the feasibility of load following operation being considered for SMR was also evaluated. In addition, the comparison between experimental results and system code analyses enhanced the reliability of system code modeling and established a foundation for the analysis of transient behaviors that are challenging to try in experiment. This research validates the natural circulation operational performance of integrated SMR designs like NuScale concept and extend confidence in next-generation SMR options, such as load-following capabilities.

Based on large PWR and SMART SMR (Small Modular Reactor) technologies, the innovative SMR,referred to as the i-SMR, is under development. The i-SMR incorporates an in-vessel helical coil steam generator.
Boiling in helical tubes differs from straight tubes due to the interplay of centrifugal and buoyancy
forces, which influence pressure drop and flow regimes. Heat transfer models used in safety analysis codes are largely based on straight-tube boiling phenomena, offering limited applicability to helical tube heat transfer predictions. Furthermore, flow instabilities under extreme conditions may compromise the structural integrity of pipes due to vibration, affecting flow stability, flow rate, and heat transfer performance, thereby potentially impacting system safety. To optimize performance and validate heat transfer in helical coil steam generators under conditions similar to the i-SMR, heat transfer tests are being conducted, and a test database is being developed. Additionally, to ensure the safety of integral reactor designs like the i-SMR, ongoing assessments aim to refine and evaluate heat transfer and pressure drop models specific to helical coil steam generators, which significantly differ in geometry and heat transfer characteristics compared to large-scale nuclear steam generators.
This study serves as an initial step in validating the heat transfer performance of the helical coil steam generators for the i-SMR. It focuses on evaluating existing heat transfer models, based on prior research, to determine their applicability to helical coil steam generators.

The passive auxiliary feedwater system (PAFS) is one of the advanced safety systems in the innovative Small Modular Reactor (i-SMR). The PAFS has a heat exchanger tube bundle submerged in the emergency cooling tank (ECT). In the PAFS, the heat is removed by the condensation in the heat exchanger tube having 3 degree inclination. To know the heat removal performance of the PAFS, the heat transfer rate for the condensation in tube should be accurately predicted. In this study, to evaluate the predictability of SPACE code for the condensation in the PAFS, the SPACE code analyses were conducted for the PASCAL and PICON experiments. For the PASCAL experiments which simulated a PAFS heat exchanger tube and the PICON experiments which simulated the condensation in the nearly horizontal tube, the heat transfer coefficient and flow regime were compared between the experimental results and the SPACE code analysis results. The SPACE code predicted well the condensation heat transfer rate in the PASCAL and PICON experiments when the condensation model developed by Ahn et al. (2014) was applied.

Supercritical carbon dioxide (sCO₂), nitrogen (N₂), and helium (He) Brayton cycles are promising power conversion systems for advanced nuclear reactors, including molten salt reactors (MSRs), sodium-cooled fast reactors (SFRs), and gas-cooled reactors (GCRs). Recuperators play a crucial role in enhancing thermal efficiency in these cycles but require volume minimization due to their high thermal duty and large size. This study investigates the thermal-hydraulic performance of straight-channel recuperators for sCO₂, N₂, and He Brayton cycles. Gas-to-gas experiments were conducted using printed circuit heat exchangers (PCHEs), covering a wide range of Reynolds (Re) numbers from laminar to turbulent regimes to accommodate various design conditions. Experimental results were analyzed based on Re numbers, and existing thermal-hydraulic correlations were evaluated for their applicability in recuperator design. In laminar regime, the developing flow effects are important for heat transfer and pressure drops. In transition and turbulent regimes, existing correlations have enough predicting performances. With the evaluated correlations, a validated one-dimensional (1-D) in-house PCHE design code was employed to determine the optimal recuperator volume while satisfying target effectiveness and pressure drop constraints. The optimal design results, derived under fixed thermal duty and pressure drop conditions, were examined across different Brayton cycle working fluids. The findings provide insights into the thermal-hydraulic performance of straight PCHE channels across a broad Re number range and offer valuable design-level guidance for recuperators in gas Brayton cycles. It is worth noting that these results contribute to improving the efficiency and feasibility of compact recuperators for advanced nuclear power systems.