Conference Agenda

Helical cruciform fuel (HCF), a novel nuclear fuel design, shows potential for enhancing power output and extending the service life of light water reactors (LWRs). While thermal-hydraulic studies on HCF assemblies are common, coupled analyses with high-fidelity neutron physics remain limited. This study establishes a CFD model of a 3×3 HCF assembly, integrating volumetric heat sources derived from neutron physics calculations to investigate flow and heat transfer phenomena.Key findings include helical variations in heat flux density (q) and wall temperature (T_w) along the flow direction. Due to the gap effect, q is lower in valley regions compared to blade regions, while T_w shows the opposite trend. Bulk temperature (T_l), however, lacks noticeable helical patterns. Under different heat source conditions with identical total power, peak values and positions of q, T_w, and T_l vary significantly. Condition one results in 70% higher q, an 18.8 K rise in T_w, and an 8 K increase in T_l compared to condition Two, with peaks occurring in different axial regions. Conversely, condition Two shows minimal axial q variation, with T_w and T_l peaking at the outlet. These results suggest a higher likelihood of boiling crises under condition one. Increased local axial power exacerbates circumferential non-uniformity in q and T_w, with heat transfer deteriorating at blade regions aligned with Twist angles of multiples of 90°, marked by reduced q and elevated T_w and T_l.

Fuel elements play a vital role in the safety and economic efficiency of nuclear power plants as one of the core components of a reactor. In addition to facilitating inter-channnel mixing between rods and enhances heat transfer, the wire-wrapped rod bundle is free from the supporting of spacer grid due to its self-locating structure through the contact between adjacent rods, which making it popular in recent research on reactor structural design. However, complex mechanical interactions often occur in wire-wrapped fuel rods under the constraints of reactor irradiation and high temperature, leading to stress concentration at the contact points of adjacent rods. This can easily cause fatigue damage to the fuel clad and affect its integrity. This study establishes a finite element model of wire-wrapped fuel rods by using the Ansys Workbench, taking into account the effects of irradiation, high temperature, and the geometric structure of the wire. The mechanical interaction characteristics of wire-wrapped rods under complex working conditions are investigated. The results obtained from this study on the mechanical characteristics of wire-wrapped rods can provide insights for the structural optimization design of fuel rods.

Externally cooled dead-end pipes, thermally and hydraulically connected to a hot source, exhibit complex physics, primarily driven by the interaction between penetrating swirl at the open end and stratified flow near the closed end of the pipe. These inherent instabilities in practice can lead to temperature fluctuations, potentially causing thermal fatigue and leakage in stainless-steel pipes. Such uninsulated pipes may be found, for example, in some 2 loop Westinghouse PVRs, where the safety injection (SI) pipes are connected directly to Reactor pressure Vessel (RPV). To better understand the thermal-hydraulic behaviour of this SI pipe configuration, CFD simulations were conducted. Despite advancements in computational power, such industry-level simulations remain challenging due to numerous uncertainties affecting prediction accuracy. Our study highlights that prediction of the turbulent swirl plus competing with the natural circulation is highly sensitive to CFD model settings, including boundary conditions, mesh resolution, turbulence models, and numerical methods (e.g., discretization schemes, solver types). These sensitivities suggest that the phenomena are unstable and chaotic. External air cooling, which induces flow stratification, emerges as the primary source of uncertainty, while unknown geometry at the inner edge of the DVI nozzle where the swirl forms, adds further complexity. Additionally, the large geometrical model restricts a systematic mesh sensitivity study, and the lack of sufficient experimental data limits the validation of turbulence modelling. All the above-mentioned aspects will be systematically reviewed in our paper, and supported by the CFD examples.

Integral effect test data obtained from the High Temperature Test Facility (HTTF) are being used for benchmarking CFD and system codes in the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data. Five system codes SAM, RELAP5-3D, GAMMA+, SPECTRA, ARIANT, and CATHARE are used to model benchmark Problem 3 Exercises 1C and 1D, which simulate the steady state and pressurized conduction cooldown (PCC) transient of HTTF Test PG-27. This test examines the PCC phenomena progression in an integral test facility scaled to the General Atomics MHTGR design. The proposed exercises include well defined boundary conditions and assumptions so that code-to-code comparisons will help identify differences between modeling approaches, numerical methods, and uncertainties in the solutions of different codes. Preliminary assessment of the results shows that in steady-state operating condition, the codes agree very well for key parameters such as coolant temperature, solid temperature and flow distribution in core regions. In PCC transient, the agreement is reasonably good. The codes predict that natural circulation is several orders of magnitude lower than steady-state flow rate and core-wise heat transfer is therefore dominated by thermal conduction and radiation. The codes also predict similar temperature trends in the solid structures but there are discrepancies in the transient behavior. This is not surprising considering the vastly different modeling schemes of a very complex core geometry.

Extreme accidental scenarios in nuclear power plants (NPPs) may involve hydrogen formation and its pressurized release into the containment building, potentially leading to unintended explosions. A fundamental understanding of the complex physical mechanisms associated with such scenarios is critical for their prevention and mitigation. This includes investigating hydrogen jet dynamics during the initial stages of release under high-pressure conditions, which are relevant for hydrogen storage systems in nuclear facilities.

This work uses high-fidelity 3-D numerical simulations based on the Implicit Large Eddy Simulation (ILES) technique to investigate the turbulent characteristics and mixing of underexpanded jets, varying initial pressure ratios, and jet diameters. First, nitrogen jets released into atmospheric nitrogen are investigated, examining pressure ratios of 60, 30, 15, and 7.5 for a 3 mm diameter jet. This serves as a simpler case to analyze the jet flow dynamics. Second, we focus on high-pressure hydrogen jets released into air with the same pressure ratios and different jet diameters of 1.5, 3, and 6 mm to represent a reactor-scale problem. Both cases are validated against experimental data with an excellent agreement.

Key findings include the influence of pressure ratio and jet diameter on the turbulent jet self-similarity and mixing shear layer dynamics. A larger jet diameter enhances self-similarity, while a decrease in pressure ratio disrupts it. Higher pressure ratios result in thicker shear layers and broader temperature ranges. These insights contribute to enhancing safety procedures and protocols in nuclear systems and other high-pressure hydrogen storage applications.