Conference Agenda

With the advantages of high energy density, easy transportation and no pollution, hydrogen is a potential energy source for large-scale application to achieve near-zero carbon emissions, as well as an important industrial raw material. Since nuclear energy could provide stable and based loaded power with zero-carbon emissions, it is an ideal primary energy to produce hydrogen that is a kind of secondary energy. Currently, High Temperature Gas-cooled Reactor (HTGR) and water-cooled Small Module Reactor (SMR) could both be used for hydrogen production by steam reforming technology. In this work, a one-dimensional reaction flow model of the reformer tube, which is the core equipment in this methane-steam reforming hydrogen production using HTGR, was developed. The simulation results were compared to the latest experimental results, demonstrating the good validation. A comprehensive parametric sensitivity analysis on the reformer tube was performed using this model, providing a useful model to analyze and design a reformer tube for hydrogen production using HTGR. Additionally, a hydrogen production system with SMR coupled with methanol steam reforming was designed using SMR as the heat supply source. Parameter sensitivity analysis of this system was performed, and the optimal reaction conditions as well as the optimal reaction parameters were determined to provide guidance for nuclear hydrogen production system design. This work provides a general concept for nuclear hydrogen production by steam reforming technology.

An advanced concept of the Passive Molten Fast Salt Reactor (PMFR) has been recently proposed in the Republic of Korea as part of efforts to develop molten salt small modular reactors. Molten salt reactor (MSR) technologies have gained attention for their improved efficiency, enhanced safety, and capability for high-temperature operation, enabling non-electric process heat applications such as hydrogen production. A key innovation of the PMFR is the natural circulation of liquid fuel salt within the reactor loop, eliminating the need for pumps. This design improves safety by reducing reactor risks associated with pump reliability. Additionally, the PMFR is designed to integrate a compact and high-efficiency supercritical CO₂ (SCO₂) power conversion system. This study evaluates the feasibility and performance of hydrogen production systems coupled with the PMFR. For power generation, the study incorporates an SCO₂ Brayton cycle, recognized for its compact size and efficiency, and models its performance to optimize the use of the PMFR's thermal output. Potential hydrogen production methods analyzed include alkaline water electrolysis (AWE), polymer electrolyte membrane (PEM) electrolysis, and high-temperature steam electrolysis (HTSE). Thermodynamic models are developed for each production method to assess their integration with the PMFR's thermal and electrical outputs. Comparative analysis reveals that HTSE outperforms other methods in terms of efficiency and compatibility with the PMFR's high-temperature operation. The findings highlight the advantages of combining advanced nuclear reactor systems like the PMFR with HTSE for sustainable and efficient hydrogen production, offering valuable insights into future energy system designs.

This study presents an integrated solar and affordable, robust, compact (ARC) fusion reactor-driven integrated energy system for the production of electricity, freshwater, and hydrogen. The main aim of the study is to develop and evaluate non-electrical applications of the ARC fusion reactor integrated energy systems. Within the scope of this study, the integrated system consists of five subsystems, including an ARC fusion reactor, a concentrated solar power system, an open feedwater Rankine cycle, a multi-effect desalination system, and a cobalt-chlorine (Co-Cl) thermochemical cycle. The analyses of each subsystem and the overall system are assessed with the approaches of energy and exergy using the first and second laws of thermodynamics. The overall efficiencies of the integrated energy system are compared with the efficiencies of the original ARC fusion reactor design. Moreover, the Shomate heat capacity equation is employed while the calculations of the Co-Cl thermochemical cycle are carried out. The energy and exergy efficiencies of each subsystem are calculated. Consequently, the integrated energy system produces approximately 129.7 MW of electricity, 2040.7 tons/h of freshwater, and 1 mol of hydrogen per second, with 44.57% overall energy and 47.91% overall exergy efficiencies.

Space nuclear reactor systems utilizing heat pipes, which can effectively transfer heat without the need for pumps, have attracted significant attention as a viable solution for lunar applications. Korea Atomic Energy Research Institute (KAERI) has designed a sodium heat pipe incorporating a braided wick to enable flexibility in bending and using sodium as the working fluid. In this study, a test assembly consisting of six sodium heat pipes, each with a diameter of 1/2 inch and a length of 1 meter, was fabricated. The 25 cm evaporator section at the bottom, simulating a reactor core, was constructed using a graphite block and electric heaters within a helium chamber. The 50 cm adiabatic section was constructed using ceramic board insulation to encase each individual heat pipe, along with an insulation box to cover the adiabatic section of the entire assembly. To ensure precise cooling and heat transfer quantification for each heat pipe, the 25 cm condenser section was designed with a dual-cooling system comprising air and water cooling jackets. The thermal performance evaluation is conducted at temperatures exceeding 700°C, with the six heat pipes collectively transferring a total heat load of 3 kW. Experimental data obtained from this test will serve as a basis for validating the design codes of lunar nuclear reactor systems utilizing heat pipes.

In this study, thermal response and ablation tests of a containment system for nuclear batteries under re-entry aerothermal conditions were conducted using a 0.4-megawatt plasma jet test facility in Jeonbuk National University. Nuclear-powered thermoelectric generators have been utilized in space due to their ability to produce heat and electricity over extended periods through radioactive fuel decay, independent of solar flux. For the safe design of space nuclear reactors and radioisotope generators, the containment system must maintain its integrity around the radioactive heat sources even in the event of an accident. In the case of an atmospheric re-entry scenario, the containment system may fail due to exposure to the high-temperature atmosphere. Therefore, carbon-based thermal protection systems are attached to the containment system for nuclear-powered thermoelectric generators. According to a case study on re-entry conditions for nuclear batteries, the peak heat flux reaches 3.4 MW/m² with a recovery enthalpy of 11.4 MJ/kg. In this study, tests were conducted under conditions of a heat flux of 7.7 MW/m² and a recovery enthalpy of 13.9 MJ/kg. Test results showed that for a 20mm diameter carbon-carbon hemispherical sample, the ablation rate and surface temperature reached 0.04 mm/sec and 1800°C, respectively, over 120 seconds. This test data can serve as a critical database for developing an evaluation model for carbon-carbon thermal protection structures for nuclear-powered thermoelectric generators in space.

Thermionic reactors, with proven success in space applications and superior power scalability, present a promising technological solution for space nuclear power systems. To investigate the operational characteristics of thermionic space reactors, a system analysis code is developed based on the TOPAZ-II reactor. This code enables coupled nuclear-thermal-hydraulic-electrical calculations. The steady-state validation of the system analysis code is conducted according to the design values. To demonstrate the transient calculation capability of this code, the transient parameters during the start-up process are compared with the results of the referenced transient analysis model. The effects of cesium vapor pressure, nuclear power, and load resistance on system-level steady-state output characteristics are analyzed, and the intervals of the boundaries for optimizing system performance are determined. The results indicate that the steady-state calculation error of the developed code is less than 2.5%. The system responses during the start-up transient process agree well with the referenced values. The transient calculation of the system analysis code with comprehensive models is more consistent with the engineering practice. When the single-boundary variation intervals of cesium vapor pressure, nuclear power, and load resistance are (0.76 torr, 2 torr), (115 kW, 135 kW), and (0.068 Ω, 0.141 Ω) respectively, the system can achieve more efficient output electrical power than that in the benchmark steady-state. These findings provide valuable insights for improving the operational strategies of thermionic space reactors, and the system analysis code could serve as a theoretical tool for safety analysis.