Nuclear Hydrogen Production System Research Articles

Integrated scheme analysis and thermodynamic performance study of advanced nuclear-driven hydrogen-electricity co-production systems with iodine-sulfur cycle and combined cycle

A promising method for massive clean hydrogen production is the Very High Temperature Reactor (VHTR)-driven Nuclear Hydrogen Production (NHP) system using the Iodine–Sulfur (IS) cycle. Research on integrated scheme and thermodynamic performance of the VHTR-driven NHP system using the IS cycle has, however, received little attention up to this point, particularly when the combined cycle is employed as the power generation cycle. In order to bridge this research gap, this work proposed and studied two distinct VHTR-driven hydrogen-electricity co-production systems with the IS cycle and combined cycle: the independent operating system and the coupled operating system. Thermodynamics was used to model these two systems, and system thermodynamic performance was examined under the baseline operating conditions. A parametric study was further conducted on how two important operating parameters affected system thermodynamic performance. The primary findings indicated that the coupled operating system outperformed the independent operating system in terms of thermodynamic performance. Additionally, both the independent and coupled operating systems could produce hydrogen at the same rate of 289.8 mol/s, with net electrical power outputs of 61.07 MW and 102.7 MW, respectively, under the baseline operating conditions. Furthermore, it was discovered that a rise in the mass flow ratio for both operating systems would result in a notable reduction in system efficiency.

Experimental and numerical investigation of leakage and diffusion accidents based on nuclear energy hydrogen production system

The high-temperature gas-cooled reactor, as a fourth-generation nuclear power system, possesses a distinct advantage of high core outlet temperature, making it a viable option for large-scale, low-carbon hydrogen production. However, before this technology can be widely adopted, it is crucial to conduct a safety assessment of the nuclear hydrogen production system, considering the unique physical and chemical properties of hydrogen. This study aimed to address this requirement by implementing two experimental setups, utilizing helium as a substitute for hydrogen. Concentration distribution measurements were performed in both open spaces and inside a wind tunnel. The concentration decay rates were found to be 0.177 under windless conditions and 0.313 in windy conditions. Subsequently, numerical models were employed to validate the feasibility of leakage and diffusion calculations for subsonic leakage jets and high-pressure underexpanded leakage jets, respectively. Moreover, multiple simulation calculations were conducted to design leakage accident scenarios based on the nuclear hydrogen production system. The corresponding concentration decay patterns were summarized, and evaluation relational expressions were proposed to determine diffusion distance and explosion distance for the nuclear energy hydrogen production system, accounting for the influence of buoyancy effects and incorporating the average concentration of combustible hydrogen clouds. Lastly, considering the worst-case accident conditions of the current system, a fan reverse blowing scheme was suggested to mitigate safety risks. The findings indicated that this scheme could potentially reduce the separation distance from 479 m to 183 m. Overall, this research contributes valuable insights into the safety assessment and design of nuclear hydrogen production systems.

Influence of obstacle morphology on safety of nuclear hydrogen production system

The safety of the nuclear hydrogen production system is a key issue that cannot be ignored in the future commercialization. The reasonable arrangement of obstacles can effectively improve the safety of the system under accident conditions. In this paper, based on the high-pressure hydrogen storage tank model of a nuclear hydrogenation plant, two mechanisms by which obstacles can effectively improve safety are analyzed. In order to determine the obstacle design scheme with outstanding effect, the response surface methodology is used to study the influence of the structural parameters and spatial position parameters of obstacles of different shapes, and the comprehensive performance of different obstacles is compared by the TNO multi-energy method. The results show that the semi-cylindrical surface can control the separation distance within 141 m at a lower processing cost, which has good engineering utilization value. This study can provide a valuable reference for the design of nuclear hydrogen production systems.

Safety analysis of leakage in a nuclear hydrogen production system

Due to its unique advantages, such as clean and pollution-free, hydrogen energy has gradually improved its energy transition position. Constructing nuclear hydrogen production systems is a necessary means to achieve large-scale hydrogen production, and the study of hydrogen leakage and diffusion behavior is critical to commercializing hydrogen production systems. In engineering practice, the distance between the hydrogen storage device and the nuclear power plant is an important indicator to measure the safety of nuclear hydrogen production. To study the influence of gas storage tank's own conditions and external environmental conditions on leakage diffusion, influencing factors such as wind speed, leakage direction, leakage diameter, leakage height, and leakage angle are discussed in the present study. By calculating severe working conditions combined with the above multiple factors, the longest distance of hydrogen diffusion is determined. Finally, peak overpressure impact generated by hydrogen explosion was evaluated, and the minimum separation distance required to avoid safety risks was predicted. The results demonstrate that when the wind direction is consistent with the leakage direction, and the leakage angle is 0°, the higher the wind speed, the larger the leakage diameter and the lower the leakage height, resulting in a longer diffusion distance. Under more extreme and severe working conditions, the diffusion distance of combustible hydrogen cloud can reach as far as 237 m. Once hydrogen diffusion explodes, the minimum separation distance required is about 338 m. This research provides an effective method for safety risk assessment of a nuclear hydrogen production system.

Thermodynamic analysis and optimization of the combined supercritical carbon dioxide Brayton cycle and organic Rankine cycle‐based nuclear hydrogen production system

The Supercritical Carbon Dioxide Brayton Cycle (SCBC) with high efficiency is a very promising thermodynamic cycle to replace the Steam Rankine Cycle (SRC) as the power cycle of a Nuclear Hydrogen Production (NHP) system. However, detailed research on the NHP system using SCBC as a power cycle has not been carried out so far and the system's thermodynamic characteristics are not well known. To fill this research gap, three promising very high-temperature gas-cooled reactor and copper-chlorine cycle-based NHP systems using different power cycles including SRC, SCBC, and combined SCBC and Organic Rankine Cycle (ORC) are proposed and studied in this work. The energy and exergy analysis method and the particle-swarm optimization algorithm have been used respectively to model and optimize the system. In addition, a parametric analysis based on two typical reactor concepts has been performed to comprehensively investigate the system's thermodynamic characteristics. Finally, a performance comparison of three systems under different hydrogen production loads has been conducted. Major results show that under different hydrogen production loads, the variations in the thermodynamic performance of the SCBC-based NHP system are sensitive to the reactor inlet temperature. Under low reactor inlet temperatures (≤587°C), the thermal efficiency of the SCBC is less than 37%. Besides, increasing the compressor pressure ratio of SCBC can enhance the system's thermodynamic performance, and the system's thermal and exergy efficiencies are improved by about 0.7% to 3.7% and 1.0% to 5.3% respectively by using ORC. Highlights Three VHTR and Cu-Cl cycle-based nuclear hydrogen production systems are studied. Three systems are modeled based on the energy and exergy analysis method. Parametric analysis and optimization under two reactor concepts are conducted. SCBC's thermal efficiency under low reactor inlet temperatures is less than 37%. System overall thermal efficiency is improved by about 0.7% to 3.7% through using ORC.

Thermo-economic analysis and optimization of the very high temperature gas-cooled reactor-based nuclear hydrogen production system using copper-chlorine cycle

An improved very high temperature gas-cooled reactor (VHTR) and copper-chlorine (Cu–Cl) cycle-based nuclear hydrogen production system is proposed and investigated in this paper, in order to reveal the unknown thermo-economic characteristics of the system under variable operating conditions. Energy, exergy and economic analysis method and particle swarm optimization algorithm are used to model and optimize the system, respectively. Parametric analysis of the effects of several key operating parameters on the system performance is conducted, and energy loss, exergy loss, and investment cost distributions of the system are discussed under three typical production modes. Results show that increasing the reactor subsystem pressure ratio can enhance the system's thermo-economic performance, and the total efficiencies and cost of producing compressed hydrogen from nuclear energy are respectively lower and higher than that of generating electricity. When the system operates at the maximum hydrogen production rate of 403.1 mol/s, the system's net electrical power output, thermal efficiency, exergy efficiency, and specific energy cost are found to be 38.77 MW, 39.3%, 41.26%, and 0.0731 $/kW·h, respectively. And when the system's hydrogen production load equals to 0, these values are respectively calculated to be 177.25 MW, 50.64%, 53.29%, and 0.0268 $/kW·h. In addition, more than 90% of the system's total energy losses are caused by condenser and Cu–Cl cycle, and about 50–60% of the system's total exergy destructions occur in VHTR. About 60% and 30% of the system's specific energy cost are respectively caused by the equipment investment and the system operation & maintenance, and the investment costs of VHTR and Cu–Cl plant are the system's main capital investment sources.

Genetic-algorithm-based fault detection of the heat transfer process in nuclear-based hydrogen production based ON Cu–Cl cycle

In the heat transfer process of nuclear hydrogen production system based on a 4-step Cu–Cl cycle, there are three major varying temperatures, namely room temperature, hydrolysis temperature and oxygen decomposition temperature. The heat requirement of the system will change with the temperatures. Therefore, it is important to identify the optimal ranges of the varying temperatures to make sure the heat requirement will be satisfied by the heat from a nuclear power plant, when the temperatures fluctuating in the optimal ranges. Genetic-algorithm-based Monte Carlo simulation method is developed in this paper to identify the optimal ranges. This method is able to obtain the optimal ranges through random sampling. The final result indicates that GA-based MCS method can be applied to identify the optimal ranges of the varying temperatures in the heat transfer process, and the ranges are highly trustable.

초고온가스로를 이용한 원자력수소생산 기술개발

미래에너지의 해법으로 원자력에너지를 이용한 물분해 수소생산시스템의 핵심기술을 개발하였다. 안전성을 보장할 수 있는 제4세대 원자로인 초고온가스로의 고열을 이용하여 황요오드 열화학적인 방법으로 물을 분해하여 수소를 생산하는 기술이다. 원자력수소생산 핵심기술은 초고온에서의 열을 공급하는 것을 모사하는 초고온 실험기술, 초고온가스로의 안전성을 모사하는 연구, 초고온가스로의 노심과 안전성을 해석할 수 있는 도구의 개발, 초고온가스로에 사용하는 연료제조기술, 물을 분해하여 열화학적인 방법으로 수소를 생산하는 기술로 구성된다. 원자력수소생산에 필요한 핵심기술을 개발하고 실험실 규모로 입증하였으며, 대규모 실용화를 위해서 선결되어할 미완성 기술을 제시하였다. 본 기술은 제4세대 원자로개발 국제공동연구로 수행한 기술로서 향후 미래의 원자로 기술이다.

Safety design consideration for HTGR coupling with hydrogen production plant

Safety requirements and design considerations are examined for a nuclear hydrogen production system that consists of High Temperature Gas-cooled Reactor (HTGR) and a hydrogen production plant by thermochemical water splitting iodine–sulfur process (IS process). Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified in order to take into account a fundamental difference in safety philosophy between the nuclear facility and chemical plant and meet requests from the potential users of nuclear heat. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process (IS plant) are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results for a reference plant showed that safe distance determined by the chemical plant regulation against combustible gas and hazardous chemical leakages comply with the plant layout design. Furthermore, the results demonstrated the feasibility of IS plant construction under non-nuclear regulation by showing that the tritium concentration in IS plant can be maintained below the regulation limit and reactor normal operation can be achieved during abnormal conditions in the IS plant. These results clarified that design considerations suggested for coupling the IS plant to HTGR are reasonably practicable. The proposed criteria can be used not only for coupling hydrogen production plants but also for other chemical plants such as steam reforming plants, etc.

Materials development for process heat exchanger (PHE) in nuclear hydrogen production system

The decomposed sulfuric acid gas (SO2/SO3/H2O) channels made of super alloys (Alloy HX in this work) in process heat exchanger (PHE) are subjected to an extremely severe corrosion environment because its operation temperature is more than 900°C. For the corrosion protection, Alloy HX was coated with SiC film in this work. The bonding between two dissimilar materials is often problematic, particularly in coating metals with ceramics protective layer. A strong bonding between SiC/Alloy HX was achieved by mixing the atoms at the interface by ion-beam: The film was not peeled-off at ≥900°C, confirming excellent adhesion, although the thermal expansion coefficient of Alloy HX is about three times higher than that of SiC. Instead, the SiC film was cracked along the grain boundary of the substrate above 700°C. At ≥900°C, the film was crystallized forming islands on the substrate so that a considerable part of the substrate surface could be exposed to the corrosive environment. In order to cover the exposed areas and cracks multiple coating/IBM process has been developed. The immersion corrosion test in 80% sulfuric acid at 300°C for 100h showed the weight retain rate was gradually increased as increasing the number of the process time.

Safety Evaluation of the HTTR-IS Nuclear Hydrogen Production System

The establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct, and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification is conducted for the HTTR-IS system, a nuclear hydrogen production system by thermochemical water splitting iodine-sulfur process (IS process) utilizing the heat from the high temperature engineering test reactor (HTTR) in order to select abnormal events, which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the identified five anticipated operational occurrences (AOOs) and three accidents (ACDs) show that evaluating items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case and therefore, the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

Development of an evaluation method for the HTTR-IS nuclear hydrogen production system

A thermochemical water splitting hydrogen production system based on the iodine sulphur (IS) process is presently under development in JAEA. The hydrogen production system is to be connected to the HTTR operating test reactor in JAEA. An important development goal for the HTTR-IS system is design and construction of the IS process to the standards of a conventional chemical industrial plant in order to simplify the cost and operation of the overall nuclear hydrogen production.The present study is conducted to meet a key technology requirement to achieve the goal, that is, to establish a method to evaluate the effects of anticipated abnormal load changes in the IS process on the reactor operation. The numerical models are developed with capability to compute abnormal transient behavior of process heat exchangers in the IS process. Furthermore, interface schemes between the RELAP5 code and the models are developed to quantify the impact of IS process load change events on the reactor operation. A preliminary analysis in case of inadvertent opening of pressure control valve in the H2SO4 decomposition procedure of the HTTR-IS system has been conducted. The results of the analysis show that the abnormal load increase of the IS process due to the decrease in boiling and dew point in the H2SO4 decomposer and to the chemical equilibrium shift of the H2SO4 decomposition is successfully simulated by the chemical reaction and H2SO4–H2O aqueous system models developed. Furthermore, the mitigation effect of a steam generator installed alongside the IS process against abnormal load transient in the reactor is clarified by the developed algorithm interfacing the chemistry model calculation with the system analysis that considers the heat conduction in heat transfer tubes. It is shown that the thermochemical models and model interface schemes developed in the present study make possible the system analysis of the HTTR-IS nuclear hydrogen production during the IS process abnormal events.

Materials Development by a Surface Modification for the Sulfuric Acid Decomposer in Iodine-Sulfur(IS)cycle for Nuclear Hydrogen Production System

AbstractEfforts have been made to develop long term sustainable materials for use above 900°C in the SO3/SO2 environments utilized in Nuclear Hydrogen Production Systems. In this study, the surface of Hastelloy X was serially modified by evaporative SiC coating, ion beam mixing (IBM), additional coatings, and final ion beam hammering (IBH). Subsequent heating above 900°C results in no peeling-off of the SiC coating layer in spite of the huge difference in their coefficients of thermal expansion (CTE). It was also found that ion beam hammering (irradiation) suppresses the vacuum sublimation of the low density (∼40% of bulk) ceramic film (bulk materials begin to sublime at ∼ 750°C in a vacuum of 1.5×10−5 torr). The sublimation rate is ≤ 30% of the bulk rate after an annealing at 950 for 2 hrs but is decreased to ≤10% of the bulk rate after irradiation with 70 keV ions to a total dose of 1×1017 N ions/cm2. Further irradiation (up to 4×1017 N ions/cm2) does not further decrease the rate. Both an immersion test in 98% sulfuric acid and the potentiodynamic polarization test suggest that the surface modified Hastelloy X has a greatly prolonged life time in the corrosive sulfuric acid atmosphere, suggesting the serial surface modification process is applicable to the thermo-chemical system for the nuclear hydrogen production.

Thermal design of a laboratory—scale SO 3 decomposer for nuclear hydrogen production

Designing of a sulfur trioxide decomposer is one of the main technical challenges in the development of a nuclear hydrogen production system. Korea Atomic Energy Research Institute (KAERI) has developed a hybrid-design decomposer to withstand the severe operating conditions. In this design, the hot gas side is equipped with printed circuit-type heat exchangers (PCHE), and the process gas side is equipped with plate fin-type heat exchangers (PFHE). In this study, the sensitivity analysis is performed to estimate the effects of the various factors on the thermal design of a sulfur trioxide decomposer for its performance test in a small-scale gas loop. The principal factors affecting the decomposer's thermal design are categorized into four groups: rate constants, geometrical parameters, heat transfer parameters, and operating conditions. Our results indicate that the rate constant of the catalyzed reaction and the decomposer retention time are the most important factors dictating the thermo-chemical behavior of the sulfur trioxide decomposer. The information obtained from this analysis will serve to guide the manufacture and operation of this laboratory-scale decomposer.

A Surface Modification of Hastelloy X by a SiC Coating and an Ion Beam Irradiation for a Potential use for Iodine-Sulfur Cycle in Nuclear Hydrogen Production System

ABSTRACTThe materials used for the SO3 decomposer in Iodine-Sulfur (IS) cycle for Nuclear Hydrogen Production System require excellent mechanical properties as well as a high corrosion resistance in SO2/SO3 environment at an elevated temperature up to 950°C. So far, no metallic materials have been suggested to be useful in such an environment. A surface modification of Hastelloy X by a SiC coating processed by an electron beam evaporative deposition has been studied in combination with an ion beam mixing (IBM) and an ion beam hammering (IBH). The simply deposited SiC film on the Hastelloy X substrate is easily peeled-off during an annealing at a high temperature due to a huge difference in their coefficients of thermal expansion (CTE), however the SiC coating on Hastelloy X prepared with IBM is sustained above 900 °C when the heating rate is less than 10 °C/min. The process of coating and IBM consists of a thin SiC film deposition, a subsequent N ion beam bombardment, and then an additional deposition of the film to the designed thickness. IBM plays a role of fastening the SiC film on the Hastelloy X substrate until the interfacial reaction takes place. Once the reaction takes place, new phases are developed at the interface under the consumption of the film and the substrate materials, producing a buffer layer. Without IBM, the SiC film tends to be easily detached during an annealing before the interfacial reaction initiates. The SiC film prepared with IBM requires a post-deposition annealing in vacuum for the interfacial reaction. However, the sublimation of SiC film prepared by an electron beam evaporative deposition occurs at the temperature above 900 °Cdecreasing the thickness of the deposited film. The sublimation of the SiC film can be prevented by IBH in which ion beams are bombarded onto the deposited film. This may be attributed to an ion beam bombardment induced densification of the deposited film. The resultant SiC coated Hastelloy X prepared by IBM and IBH exhibits a high corrosion resistance in a sulfuric acid at 300 °C, suggesting a possible application for the IS cycle.

Safety Assessment of VHTR Hydrogen Production System against Fire, Explosion and Acute Toxicity

The Japan Atomic Energy Agency has been developing a nuclear hydrogen production system by using heat from the Very High Temperature Reactor (VHTR). This system will handle a large amount of combustible gas and toxic gas. The risk from fire, explosion and acute toxic exposure caused by an accident involving chemical material release in a hydrogen production system is assessed. It is important to ensure the safety of the nuclear plant, and the risks for public health should be sufficiently small. This report provides the basic policy for the safety evaluation in cases of accident involving fire, explosion and toxic material release in a hydrogen production system. Preliminary safety analysis of a commercial-sized VHTR hydrogen production system, GTHTR300C, is performed. This analysis provides us with useful information on the separation distance between a nuclear plant and a hydrogen production system and a prospect that an accident in a hydrogen production system does not significantly increase the risks of the public.

Development of a Nuclear Hydrogen Production System by Dimethyl Ether (DME) Steam Reforming and Related Technology

Targeting a hydrogen production system using heat produced by a nuclear reactor at about 300°C, we are developing a dimethyl ether (DME) steam reformer and hydrogen purification systems as well as catalysts for DME reforming. The use of heat from a nuclear reactor suppresses the CO2 concentration change in the atmosphere. In our developments, a catalyst, consisting of mixed oxides, produced hydrogen at a rate of about 1.9 Nm3/h per catalyst volume (m3) at about 300°C. Subsequently, the DME steam reformer achieved a hydrogen production rate of approximately, at least, 1.4 Nm3/h at about 300°C, by absorbing heat from the supplied steam. The aforementioned hydrogen production system via DME steam reforming is to be demonstrated using a thermal power plant. DME steam reforming by using waste heat and the utilization of the produced hydrogen within a combined cycle power plant can reduce fuel consumption, for instance, by about 17% compared to the case of direct DME combustion. The total system, with the use of DME, was compared with the methane case. If necessary, the byproduced CO2 may be injected into coal seams, increasing CH4 production via the substitution of CO2 for CH4 on coal, where CO2 adsorption is expected to be stronger than the CH4 adsorption.

Thermodynamic Analysis of Thermal Efficiency of HTR-10 Hydrogen Production System

This paper presents thermodynamic analysis of the thermal efficiency of the 10 MW high temperature gas cooled reactor (HTR-10) hydrogen production system. The global reaction for the equilibrium reaction model is introduced. An analytical expression for the thermal efficiency is developed using the global reaction. For the specified temperature and pressure the thermal efficiency can be computed with the solution of the equilibrium. The investigation provides a more realistic limit for the efficiency of the nuclear hydrogen production system. The influence of the temperature, latent heat, steam-to-carbon ratio and pressure on the thermal efficiency is analyzed. Varying the temperature there is a maximum thermal efficiency for the specified pressure and steam-to-carbon ratio. The latent heat influences the thermal efficiency significantly, especially at the high temperature condition. Also varying the steam-to-carbon ratio there is a maximum thermal efficiency for the specified pressure and temperature. The process should be operated with high steam-to-carbon ratio to obtain maximum thermal efficiency when the reforming temperature is low and pressure is high. The maximum value is 68.9% within the range of the pressure greater than 1 MPa and steam-to-carbon ratio greater than 2. Comparison of theoretical results to experimental data is carried out.