Conference Agenda

The artery alkali-metal heat pipes in reactors are essential for energy transfer, with dynamic thermal performance, such as two-phase circulation startup and capillary heat transfer limits, posing challenges to overall reactor performance. Investigating the capillary dynamics behind the transient thermal load is crucial for understanding the operational characteristics of artery heat pipes. This work aims to investigate the complex heat and mass transfer phenomenon of the capillary limit, which is characterized by dynamic non-equilibrium and multi-scale and multi-physical coupling, by conducting this research from the three dimensions of micro-mesoscopic mechanism, macro heat transfer characteristics, and reactor system operating patterns. By developing a theoretical framework for dynamic capillary heat transfer, a dynamic thermal analysis model for the artery heat pipes has been established and validated through experiments. The average error of the capillary dynamics model compared to experiments is 3%, while the dynamic heat transfer model shows less than 10% error against CFD simulations and under 20℃ error compared to steady-state and transient experimental results, confirming the model's accuracy. Additionally, the study investigates the correlation between capillary dynamics and dynamic heat transfer phenomena, identifying three startup phases of free molecular flow, continuous flow expansion, and continuous flow. It categorizes capillary limits into gas-phase and liquid resistance-dominated types based on two-phase countercurrent circulation. By combining the weak feedback characteristics of fast reactors with the dynamic heat transfer of artery heat pipes, the study proposes operational strategies for typical heat pipe reactors, examining system behavior during startup and under transient thermal loads.

The focus is on a 10 MW thermal Advanced High Temperature gas-cooled Micro Reactor (AMR) currently being designed. The reactor will employ prismatic graphite blocks for structural and moderator material. There will be 420 fuel assemblies in the core using low enriched TRISO fuel contained in borings within the fuel graphite blocks that allow annuli for cooling. The thermal-hydraulic behaviour of the reactor under steady-state conditions and during a Depressurized Loss of Forced Cooling (DLOFC) event has been simulated employing an axi-symmetric systems network model using Flownex SE. Under steady-state conditions the helium coolant enters the reactor at 320 C and exits at 750 C. It is found that the bottom of the core is 403 C hotter than the top of the core and in the radial temperature gradient is distorted due amongst others to an average drop in temperature of 220 C between the last fuel ring and the outer reflector (OR). The OR transfers 618 kW to the coolant flowing up the risers placed in the OR, preheating the coolant 346.8 C. The reactor cavity cooling system (RCCS) rejects 86.1 kW. During the first 5 seconds of the DLOFC the mass flow rate through the initially increases due to the blowdown effect, and the heat transfer to the also fluid increases initially. Subsequently the heat rejected by the RCCS reach a maximum of 109 kW. It found that the heat released by the solids can constitute up to 45.5% of the heat rejected by the RCCS.

The eVinci™ Microreactor which is under development by Westinghouse Electric Company could bring a cost-competitive and reliable nuclear energy source to the world. The small size of the eVinci microreactor allows for transportability and rapid, on-site deployment.

Instead of a fluid-based primary coolant system normally seen in nuclear power plants, eVinci Microreactor adopts heat pipes to transfer heat from the reactor to the Primary Heat Exchanger (PHX). The heat pipe design enables passive core heat removal which eliminates numerous components needed in active coolant systems and makes the eVinci microreactor a pseudo “solid-state” reactor with minimal moving parts.

The eVinci Nuclear Test Reactor (NTR) is a nuclear test facility dedicated for eVinci microreactor’s development. The NTR will provide critical engineering information for analysis code validation to support commercial licensing. A CFD model has been developed to support NTR PHX design optimization. A two-step method was employed for the NTR PHX CFD modelling. Step 1: A series of cases for single heat pipe finned-sleeve tube were simulated with the finned channel simplified as a porous media. The expressions for resistance and heat transfer coefficient were derived for porous media. The results were benchmarked to the test data. Step 2: Applied the derived expressions for porous media parameters from Step 1 to a full PHX CFD model. The results from Step 2 were used to help PHX design optimization. In this paper the two steps of the NTR PHX CFD model development were presented.

In this paper a multiphysics integrated full-core 3D model and the analysis results of the Westinghouse Nuclear Test Reactor (NTR) are presented, coupling the neutronic and thermal analysis in the reactor core and the heat transfer in the primary heat exchangers. The NTR reactor is an advanced 2~3 MWt transportable heat pipe cooled microreactor currently developed by Westinghouse. It is an epithermal reactor with prismatic solid core using TRISO particle fuel embedded in cylindrical fuel compacts.

The software tools used are finite element (FE) solvers developed in the framework of the Multiphysics Object Oriented Simulation Environment (MOOSE) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program sponsored by the US Department of Energy (DoE). MOOSE-based multiphysics modules of neutronics and thermal-hydraulics are coupled in solving the 3-D fission power and temperature distributions in the full core NTR reactor model. Furthermore, the reactor core model is coupled with 1-D flow models of the cooling air channels over the condenser sections of heat pipe, simulating the heat transfer in the Primary Heat Exchanger (PHX). Non-uniform flow and inlet temperature among air flow channels are informed by detailed computational fluid dynamics (CFD) calculation of the PHX. Using this integrated model, several Design Basis Accidents (DBAs) identified for the NTR design are simulated, including the accidents initiated from inadvertent control drum rotation (reactivity insertion), total loss of PHX, and single heat pipe failure.

The Single Primary Heat Extraction and Removal Emulator (SPHERE) facility at Idaho National Laboratory (INL) has been instrumental in advancing the development and validation of heat pipe technologies for microreactor applications. As a part of these efforts, long-duration heat pipe tests are required to assess long-term reliability concerns related to wick degradation, corrosion, manufacturing methods, and compatibility of materials. This paper presents the findings of a long-duration test conducted at the SPHERE facility, focusing on the performance and reliability of a high-performance, defined as over 2kW, heat pipe under sustained operational conditions. The tests emulated the anticipated common thermal characteristics of microreactor concepts. The results show the robustness of heat pipes, with a significant amount of data collected on the ratio of heat losses to heat transported and degradation rates over an extended period. Key data and performance metrics, including time series of temperatures, axial temperature profiles, thermal response times, and heat transfer capabilities, the thermal output over thermal input, were reported and discussed. These findings provide critical insights into the design and optimization of heat pipes, underscoring their potential to enhance the safety and efficiency of next-generation reactor concepts.

Heat pipes are two-phase heat transfer devices that enable passive removal of heat from the reactor core to the power conversion system in heat pipe-cooled microreactor designs. Experimental investigations of heat pipe transients are needed for technology demonstration, verification and validation of numerical codes, and the establishment of regulatory requirements. The Single Primary Heat Extraction and Removal Emulator (SPHERE) facility at Idaho National Laboratory (INL) serves as a platform for evaluating the dynamic response of high-temperature heat pipes under a variety of operating conditions. The present work details the experimental investigation of a high-performance, defined as over 2kW sodium heat pipe subjected to rapid input power fluctuations induced by sudden changes in the evaporator temperature setpoint. In addition, the heat pipe was subjected to an asymmetrical heat load where a subset of heaters operated at 30% and 70% below their nominal power. These experimental conditions were chosen to simulate thermal and operational stresses expected to be encountered in microreactors to provide data on heat pipe behavior during such important transient events. Key data and performance metrics, including time series of temperatures and strains, axial temperature profiles, thermal response times, and heat transfer capabilities, the thermal output over thermal input, were reported and discussed. The results highlight the resilience of heat pipes, revealing their potential to maintain thermal stability and efficiency under varying power loads. Lastly, the paper concludes with a discussion on the significance of the results and their implications for future research.