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1.
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.  相似文献   

2.
3.
With many advantages, hydrogen is considered as the fuel of the future. But there is no natural resource of hydrogen and it must be produced by other kinds of energy. As for the primary energy, nuclear energy is a promising alternative. Using heat from nuclear reactor to produce hydrogen is receiving more and more concerns in recent years. This paper mainly emphasizes the study of the direct contact pyrolysis (DCP) of methane using heat from nuclear reactor. A facility was designed to investigate the efficiency of DCP process in certain conditions. The experimental results show that this process produces only hydrogen and carbon. The conversion efficiency increases with temperature and residence time, but decreases as flow rate increases. The highest efficiency of DCP obtained in this exoedment is about 22%.  相似文献   

4.
Nuclear energy can provide heat and electricity to produce hydrogen. Thermo-chemical and electrical decomposition of water have been studied as a hydrogen source. In this study, the cost of hydrogen supply for transportation usage was evaluated. The total cost for a centralized hydrogen production consisted of production cost, delivery cost, and station cost. The total cost for hydrogen production using nuclear energy can be at least comparable to that of steam-methane reforming, if the cost of carbon dioxide fixation was included. The delivery cost can be reduced by optimizing the size of hydrogen production and delivery distances. The hydrogen station cost was found out to be about 50% of the hydrogen supply cost. The optimum thermal power of nuclear power plants for hydrogen production was estimated based on the cost evaluation.  相似文献   

5.
The thermochemical sulfur–iodine cycle is studied by CEA with the objective of massive hydrogen production using nuclear heat at high temperature. The challenge is to acquire by the end of 2008 the necessary decision elements, based on a scientific and validated approach, to choose the most promising way to produce hydrogen using a generation IV nuclear reactor. Amongst the thermochemical cycles, the sulfur–iodine process remains a very promising solution in matter of efficiency and cost, versus its main competitor, conventional electrolysis. The sulfur–iodine cycle is a very versatile process, which allows lot of variants for each section which can be adjusted in synergy in order to optimise the whole process. The main part of CEA's program is devoted to the study of the basic processes: new thermodynamics data acquisition, optimisation of water and iodine quantity, optimisation of temperature and pressure in each unit of the flow-sheet and survey of innovative solutions (membrane separations for instance). This program also includes optimisation of a detailed flow-sheet and studies for a hydrogen production plant (design, scale, first evaluations of safety issues and technico-economic questions). This program interacts strongly with other teams, in the framework of international collaborations (Europe, USA for instance).  相似文献   

6.
Japan Atomic Energy Research Institute carries out research and development on accelerator-driven system (ADS) to transmute minor actinides and long-lived fission products in high-level radioactive waste. The system is composed of high intensity proton accelerator, lead-bismuth spallation target and lead-bismuth cooled subcritical core with nitride fuel. About 2500 kg of minor actinide is loaded into the subcritical core. Annual transmutation amount using this system is 250 kg with 800 MW of thermal output. This transmutation amount corresponds to the amount of minor actinides produced from 10 units of 1GWe power reactors annually. A superconducting linear accelerator with the beam power of 20–30 MW is connected to drive the subcritical core. To maximize the transmutation efficiency, the nitride fuel without uranium, such as (Np, Am, Pu)N, is selected. The nitride fuel irradiated in the ADS is reprocessed by pyrochemical process followed by the re-fabrication of nitride fuel. Many research and development activities are under way and planned in the fields of subcritical core design, spallation target technology, lead-bismuth handling technology, accelerator development, and minor actinide fuel development. Especially, to study and evaluate the feasibility of the ADS from physics and engineering aspects, the transmutation experimental facility (TEF) is proposed under a framework of the High-Intensity Proton Accelerator Project.  相似文献   

7.
Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design.In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine-Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant.It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.  相似文献   

8.
Hydrogen production by high temperature electrolysis with nuclear reactor   总被引:1,自引:0,他引:1  
High Temperature Electrolysis (HTE) is a promising method because its most parts consist of environmentally sound and common materials. Hydrogen production efficiency of HTE was evaluated about the process coupling with high temperature gas cooled reactor. This process can be expected to accomplish over 53% hydrogen production efficiency at HTE operating temperature of 800 °C. As a demonstration of hydrogen production by HTE, a unit housing 15 tubular cells, where yttria-stabilized zirconia (YSZ) was used as electrolyte, was constructed, and accomplished 130 NL/h hydrogen production. In this experiment, measured hydrogen production rate has good agreement with calculated hydrogen production rate based on applied current. To design and construct large amount of hydrogen production unit, it is important to predict the thermal and electrochemical features of the unit. To predict them, the simulation technology has been developed. From the comparison between single tubular cell experimental result and simulation result, good agreement based on current–voltage characteristic was acquired.  相似文献   

9.
Several potential deployment scenarios of high-temperature gas reactor (HTGR) cogeneration systems for the simultaneous production of hydrogen and electricity have been proposed recently. They can operate under different conditions and thereby satisfy different hydrogen and electricity demand scenarios, but their performance must be studied to demonstrate thermodynamic feasibility and economic profitability to attract sufficient investment for their deployment. Therefore, this study uses exergy analysis and exergybased costing analysis methods to analyze HTGR cogeneration system performance thermodynamically and economically over the entire range of potential operating conditions by calculating performance indicators, exergy efficiency, and the specific cost per unit product. Furthermore, three optimized deployment scenarios with high performance for satisfying three different hydrogen demand scenarios are proposed based on the analysis results. The proposed three deployment scenarios show that the HTGR cogeneration system is thermodynamically efficient and economically competitive compared with other hydrogen and electricity generation systems. The feasibility of exergy analysis and exergy-based costing analysis methods for analyzing the HTGR cogeneration system so as to propose optimized deployment scenarios is demonstrated by the obtained findings practically.  相似文献   

10.
What is the future of hydrogen (H2) produced from nuclear energy? Assuming that economically competitive nuclear H2 can be produced, production of H2 may become the primary use of nuclear energy and the basis for both a nuclear-H2 renewable (solar, wind, etc.) energy economy and a nuclear-H2 transport system. The technical and economic bases for these conclusions are described. In a nuclear-H2 renewable energy economy, nuclear energy is used to produce H2 that is stored and becomes the energy-storage component of the electrical generating system. The stored H2 replaces piles of coal and tanks of liquid fuel. Capital-intensive renewable energy sources and nuclear reactors produce electricity at their full capacity. The stored H2 is used in fuel cells to produce the highly variable quantities of electricity needed to fill the gap between the electricity demand by the customer and the electricity generated by the rest of the electrical generating system. Hydrogen is also used to produce the liquid or gaseous transport fuels. This energy-system architecture is a consequence of the fundamental differences between the characteristics of electricity (movement of electrons) and those of H2 (movement of atoms). Electricity can be generated, transformed, and used economically on either a small or a large scale. However, it is difficult to generate, store, and transform H2 economically on a small scale. This distinction favors the use of large-scale nuclear systems for H2 production.  相似文献   

11.
The possibility of a hydrogen production system for Fuel Cell (FC) vehicles, which was zero carbon dioxide emission based on nuclear power, was investigated. The reactivity of calcium oxide to carbon dioxide was used for the carbon dioxide fixation and also for heat source of fuel reforming in experimental discussion. Methane was chosen as the first candidate reactant for steam reforming. Simultaneous reaction of methane reforming and carbon dioxide fixation by calcium oxide was demonstrated in a reactor packed with a reforming catalyst and calcium oxide. High-purity hydrogen, of which the concentration was higher than one at reaction equilibrium of conventional reforming, was generated from the reactor under mild operation conditions at temperature of 500–600°C and under pressure of 101 MPa. The efficiency of the fuel reforming system was estimated from the experimental results. The proposed system was expected to be applicable as a hydrogen carrier system in carbon dioxide zero-emission FC vehicles based on nuclear power.  相似文献   

12.
To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels.

The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology.

As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.  相似文献   

13.
In the present paper,we carried out a theoretical study of dielectric barrier discharge (DBD) filled with pure methane gas.The homogeneous discharge model used in this work includes a plasma chemistry unit,an electrical circuit,and the Boltzmann equation.The model was applied to the case of a sinusoidal voltage at a period frequency of 50 kHz and under a gas pressure of 600 Torr.We investigated the temporal variation of electrical and kinetic discharge parameters such as plasma and dielectric voltages,the discharge current density,electric field,deposited power density,and the species concentration.We also checked the physical model validity by comparing its results with experimental work.According to the results discussed herein,the dielectric capacitance is the parameter that has the greatest effect on the methane conversion and H2/CH4 ratio.This work enriches the knowledge for the improvement of DBD for CH4 conversion and hydrogen production.  相似文献   

14.
ABSTRACT

A new approach for generating nuclear data from experimental cross-section data is presented based on Gaussian process regression. This paper focuses on the generation of nuclear data for proton-induced nuclide production cross-sections with a nickel target. Our results provide reasonable regression curves and corresponding uncertainties and demonstrate that this approach is e?ective for generating nuclear data. Additionally, our results indicate that this approach can be applied in experimental design to reduce the uncertainty of generated nuclear data.  相似文献   

15.
One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H2O, H2SO4 and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.  相似文献   

16.
The irradiation and annealing behavior of Chinese A508-3 reactor pressure vessel (RPV) steel (0.04 wt% Cu) after 3 MeV Fe-ion irradiation ranging from 0.1 to 20 dpa at room temperature (called RTRPV) and high temperature (250?°C, called HTRPV) was studied by positron annihilation Doppler broadening (PADB) spectroscopy and nano-indentation hardness. PADB showed that the density of vacancy-type defects was higher for low-temperature irradiations. The higher hardness was found after high-temperature irradiation because of the formation of solute clusters during irradiation. Positron annihilation measurements revealed the interaction and clustering of vacancies with solute clusters which were introduced by Fe-ion irradiation. For both RTRPVs and HTRPVs, the positron defect parameter and positron diffusion length showed the recovery of the irradiation-induced defects. Total recovery was observed after annealing at 450 °C.  相似文献   

17.
在深入分析相关领域研究发展状况的基础上,提出了具有较好技术可行性的聚变高温制氢反应堆概念(称之为FDS-Ⅲ),包括具有先进等离子体物理和技术水平的聚变堆芯、先进高温锂铅包层(HTL)、可减少热流分布密度的"垂直靶板"偏滤器以及相应的功率转换系统。尤其是提出了HTL包层新概念,其特点是选用技术基础相对成熟的低活化铁素体/马氏体钢作结构材料,在锂铅流道中使用可耐高温的多层流道插件,实现约1000℃的出口温度,可应用于制氢。初步性能分析表明FDS-Ⅲ制氢堆及其包层概念具有较好的技术可行性。  相似文献   

18.
《Fusion Engineering and Design》2014,89(7-8):1223-1226
Indian LLCB – TBM uses liquid Lead-Lithium (Pb-Li) as tritium breeder, neutron multiplier and coolant. Tritium bred in liquid PbLi stream has to be recovered by tritium extraction system. Therefore, a reliable sensor with quick response time for measurement of hydrogen isotope is essential.A hydrogen isotope sensor in liquid Pb-Li, based on permeation of hydrogen isotopes through metal (sensor material) is designed. The capsule shaped sensor, made of iron membrane coated with Pd from inside (downstream side), allow hydrogen isotope to permeate through it. The design work mainly includes the selection of proper material, its thickness and surface conditions, which is to be supported by numerical calculations for optimization of maximum permeation flux, fast response and fabricability. The numerical calculation utilizes a physical model having recombination of two hydrogen isotope atoms at the surface and atomic diffusion through the bulk. In this work, design calculations based on numerical simulation and fabrication procedure of the hydrogen isotope sensor are presented.  相似文献   

19.
方勤学 《核技术》1998,21(4):251-254
作为应用基础研究,核技术的“国家目标”应当定位在江泽民主席在全国科学技术大会上指出的“为未来经济发展提供动力和成果储备”。无论从社会经济发展需求对核技术的牵引还是从核技术学科本身内在的发展规律出发,都要求核技术把与其他学科的交叉摆在首要位置上。核技术在与其他学科交叉时,要把研究的重点放在学科交叉的深层次上;要选择所交叉学科中的热点问题发挥作用;要有长远发展的战略眼光。为了推动核技术研究成果向生产转化,有必要建立符合市场规律的核技术风险投资机制。  相似文献   

20.
In this study reactor core geometrical optimization of 200 MWt Pb–Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 °C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 °C and the maximum coolant outlet temperature less than 700 °C. By taking into account of the hydrogen production as well as corrosion resulting from Pb–Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 °C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.  相似文献   

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