Conference Agenda

Accurately modeling BWR core instability phenomena using coupled system codes presents significant challenges, particularly due to numerical diffusion in the calculations, which can dampen flow and power oscillations during the analyzed scenarios. This study examines the impact of numerical diffusion in the TRACEv5p9/PARCSv3.4.3 coupled code when modeling out-of-phase neutron flux oscillations. A 3D core model of a General Electric BWR/6 reactor is developed under flow-power conditions representative of a core stability test. To induce an instability scenario, an out-of-phase oscillation aligned with the first azimuthal mode is triggered through control rod maneuvers. The thermal-hydraulic core channel layout is defined based on the dominant Lambda modes of the reactor core, ensuring consistency between the modeled flow distribution and the expected oscillation patterns.

To mitigate numerical diffusion, TRACE incorporates several second-order numerical schemes. A comparative analysis is conducted between the default first-order upwind scheme and higher-order methods to evaluate their impact on numerical accuracy. The results highlight the importance of selecting an appropriate numerical scheme to minimize diffusion effects and improve the predictive capabilities of instability behavior in BWR reactors.

There has been a growing demand for floating nuclear power plants (FNPPs) to reduce greenhouse gas emissions and provide remote energy supply in recent years. Unlike conventional land-based nuclear power plants, FNPPs experience continuous changes in heat transfer and flow characteristics due to ocean motion. In this context, the effect of ocean motion on the critical heat flux (CHF) has been studied. However, the available studies are limited, and the range often differs from the operational range of nuclear power plants. In addition, the nuclear fuel used in nuclear power plants has a cosine-shaped axial power distribution; however, experimental studies under ocean motion reflecting a cosine-shaped axial power distribution are needed.

This study conducted a flow boiling CHF experiment using the NEOUL-H platform, capable of simulating heaving motion. To simulate the ocean environment, the experiment was conducted with a period of 3-6 seconds and a maximum acceleration of 0.6 g. The CHF test loop used R134a as the working fluid, and the experiment was conducted under thermal-hydraulic conditions corresponding to PWR operating conditions through fluid-to-fluid scaling. The test section consists of a single heater rod with annular channel, and the axial power profile of the heater is cosine-shaped.

In the experiments, CHF was measured under static and heaving conditions. In the heaving condition, CHF decreased compared to the corresponding static condition. We also found that the magnitude of CHF variation depended on the thermal-hydraulic conditions, such as mass flux and pressure, and the heaving conditions, such as period and amplitude.

Three-period minimal surface (TPMS) heat exchangers have great potential in nuclear engineering because of their compact design and excellent thermal and physical properties. To further improve the performance of heat transfer capability, control factors α and β are introduced based on the standard Gyroid function to regulate the closure of the "through-hole" structure and the surface "fold" microstructure, and the flow heat transfer characteristics of the modified Gyroid TPMS structure heat exchanger are investigated based on numerical simulation and experimental measurements. The results show that with the increase of α value, the extreme temperature (Tmax) of the Gyroid structure decreases by 16.1-27.9 K, the convective heat transfer coefficient (h) increases by 28.3-33%, and the Nussel number (Nu) increases by 1.4-3.2%. With the increase of β value, the extreme temperature (Tmax) of the Gyroid structure decreased by 4.9-7.4 K, the convective heat transfer coefficient (h) increased by 4.3-8.2%, and the Nussel number (Nu) increased by 0.7-3.5%. The "through hole" closure and the addition of the surface "fold" microstructure significantly improve the convective heat transfer performance of the TPMS structure and increase the pressure drop. Taking the comprehensive performance evaluation factor (PEC) as the evaluation index, it is recommended to choose α=2.0 or β=0.8 to achieve the best effect. The scheme and structure of this study can provide a new idea for the further improvement of the TPMS structure heat exchanger.

Fluid-structure interaction (FSI) is commonly seen in nuclear power plants, such as fluid flow in piping systems and steam generator tube. While FSI analysis methods based on finite element methods (FEM) are widely used, they face difficulties in handling structural fractures and large deformations due to the complexity of mesh re-generation. Lagrangian particle methods offer a promising alternative, enabling stable computation of these phenomena. However, challenges remain, such as conservation in the discretization system which is important for stable simulation.

To address these challenges, this study presents a novel Lagrangian-Lagrangian FSI solver with a physically consistent particle method moving particle hydrodynamics (MPH-MPH). The governing equations for elastic bodies and incompressible flows are discretized using the Moving Particle Hydrodynamics (MPH) method.

Several benchmark tests, including single bar vibration, a hydrostatic water column on an elastic plate, and a dam break with an elastic gate, verified and validated the method’s accuracy. The calculation results had good agreement with theoretical predictions and experimental data.

In summary, the MPH-MPH method shows significant potential for solving FSI problems involving large deformations and fractures.

The enhancement of Latent Heat Thermal Energy Storage (LHTES) systems through fin geometry optimization remains a critical challenge for leveraging the full potential of renewable energy sources. This study focuses on optimizing the geometries of tree-shaped fins to enhance power and energy densities in LHTES systems. The goal is to find branch designs with high energy and power density through a novel surrogate model-based optimization strategy that explores a broad design space. The surrogate models applied, including linear regression, principal component analysis-based linear regression, artificial neural networks, and random forest, are evaluated for their predictive performance. The random forest model demonstrates superior accuracy in predicting targets. The optimization process results in a Pareto-optimal design with a volume fraction of 33.9%. This optimal design substantially enhances the system's power density by 61.6% compared to conventional plate fins at an equivalent energy density. This optimized design improves energy and power density, achieving a uniform end-to-branch distribution, which is a pivotal factor for consistent temperature distribution and improved thermal efficiency. By integrating surrogate-based optimization with broad ranges of the tree-shaped fin design, this research has significantly improved the operational efficiency of LHTES systems. This research promises more effective thermal management and provides a methodological framework for design innovation in thermal energy storage.

The pursuit of sustainable and clean energy has intensified global interest in nuclear fusion, particularly through tokamak reactors, which are seen as promising candidates for safe and efficient energy generation. A crucial aspect of their development is the study of the Balance of Plant (BoP). It includes all the systems belonging to the main heat transfer chain, devoted to the plasma power removal and to its conversion into electricity. This research involves BoP thermal-hydraulic analyses, focusing on the cooling system behavior, carried out by using RELAP5 system thermal-hydraulic code.

Given the pulsed nature of tokamak operation, alternating plasma pulses and dwell phases, special attention is given to transient analysis. These transient conditions pose challenges to reactor operations, influencing its efficiency, thermal stability, and safety. For this, the study includes an investigation of thermal storage systems designed to accommodate plasma power variations. A study of Intermediate Heat Transfer System (IHTS) design is proposed, which would couple the primary circuit with the Power Conversion System (PCS).

The primary goals are to evaluate the BoP response under both normal and off-normal conditions, including potential accident scenarios, to assess the plant operational efficiency and safety margins. This not only improves the understanding of energy transfer and heat management in fusion reactors but also offers information for the design and implementation of safety measures. This research contributes to optimizing tokamak design and operation, while providing pre-conceptual designs for main cooling system components. These insights and data aim to support the development of tokamak fusion reactors.