Conference Agenda

The demand for nuclear energy is rapidly increasing and, as such, the deployment of new reactors must be accelerated. Microreactors are being designed to provide electricity and heating for remote applications. Operating the microreactors in these applications will take advantage of remote and autonomous systems, which will be based on digital twin simulations. The development and validation of real-time digital twins is therefore a necessary component for microreactor deployment. The intended coupling of the digital twin control system to hardware provides opportunities to leverage methods that combine hardware and software. Hardware-in-the-Loop (HIL) testing can be integrated with a digital twin, in which an experimental subsystem will replace a region of the digital twin. As-built components can be tested directly, reducing time spent in intermediate modeling steps, and can be used for validation by providing real data for a portion of the reactor.

This paper will present the experimental setup, model development, and results of HIL testing combined with a digital twin of a microreactor cooled with heat pipes. The experimental subsystem will provide thermal hydraulic data of a hexagonal unit cell made up of a heat pipe and the surrounding region. The design of the experimental subsystem and the digital twin will be optimized to demonstrate the method and will not represent an existing microreactor. Validation of the digital twin will be performed by replacing subregions of the core with the experimental subsystem in the digital twin and comparing the results with those from the digital twin alone.

Heat pipes are passive two-phase heat transfer devices utilized in applications such as core cooling for nuclear microreactors, high-efficiency heat exchangers, and other advanced energy systems. The two-phase flow and heat transfer dynamics within heat pipes are often highly complex, particularly during transients and under vertical operating conditions. The present work develops a comprehensive heat pipe transient experimental database for a vertical heat pipe of approximately 2 meters in length using water as the working fluid, with the reported data including internal measurements of operating pressures, pressure drops, liquid film temperatures, evaporator wall temperatures, and vapor core temperatures. In particular, vapor core temperatures were obtained using a fiber optic distributed temperature sensor running along the entire heat pipe length. The database includes power input and condenser coolant flow rate transients to enable the evaluation of the heat pipe’s response to changes in both evaporator and condenser conditions. Experiments were conducted for two different wick types: annulus-screen and wrapped-screen. Important phenomena identified include vapor generation in the annulus and the presence of a subcooled liquid plug near the condenser endcap. The data obtained can be readily used for verification and validation of numerical modeling tools under development for heat pipe microreactor analysis.

Heat pipe cooled reactor (HPCR) is in the spotlight as one of the advanced reactor with the advantages of high inherent safety and compactness. HPCR was originally designed for application in space, low-efficiency power conversion systems were applied. Recently, the HPCR system for power generation has been actively studied with high efficiency power conversion system such as supercritical CO2 Brayton cycle. However, due to compact size of HPCR, there was a problem that the size of the intermediate heat exchanger that satisfies the core power increased. In this regard, we suggested new design of compact intermediate heat exchanger based on printed circuit heat exchanger (PCHE). The heat exchanger consisted of “heat pipe layers” in which heat pipes were inserted and “cooling channel layers” in which the cooling channels were machined. The structural integrity of each layer was evaluated based on ASME standard, and flow uniformity was evaluated through CFD. Based on the temperature distributions of the heater for heat pipe simulation and the heat exchanger body which were measure with fiber optic sensor (FOS), thermal contact resistance and overall thermal resistance of heat exchanger were measured. Through this study, the transportability of the designed heat exchanger was evaluated, and the possibility of comprehensive analysis through integration with the heat pipe and the reactor core model was confirmed.

Heat pipe-cooled microreactors (HPMRs) utilize alkali metal heat pipes for efficient and passive heat transfer. Simulating startup and shutdown of HPMRs require accurate modeling of transient heat pipe behavior. In this study, a 2-D transient heat pipe analysis code, SNUHTP, was developed with a melting/solidification model to simulate frozen startup and phase change effects. Transient verification against an analytical lumped model and steady-state validation using sodium heat pipe experiments showed good agreement in normal operation range of sodium heat pipe. The melting/solidification model was verified with a 1-D Stefan problem, and transient validation using the SAFE-30 heat pipe experiment showed delay in temperature rise due to latent heat effects. The results demonstrate that SNUHTP effectively predicts transient and steady-state heat pipe behavior, supporting its application to HPMR analysis.

The heat pipe-cooled reactor offers numerous advantages, including a compact design, high power density, exceptional reliability, and intrinsic safety features, making it a promising candidate for future mobile power generation systems. This reactor employs a solid-core design where high-temperature heat pipes establish a direct link between the reactor core and the energy conversion system, creating a compact and modular configuration. Despite its advantages, the intricate multiphysics interactions within the system pose considerable challenges for comprehensive analysis. To tackle this issue, this study proposes a multiphysics coupling analysis framework tailored for heat pipe-cooled reactors, developed within the OpenFOAM platform. The framework integrates neutron physics, core thermal transfer, heat pipe dynamics, and thermoelectric conversion models. Its accuracy is verified against experimental data from the KRUSTY space heat pipe reactor's ground-based nuclear testing. A complete system simulation of KRUSTY is then conducted, emphasizing the interplay of multiphysics phenomena under nuclear thermal power conditions.