Papers by Keyword: HTGR

Abstract: The primary coolant of High-Temperature Gas-cooled Reactor (HTGR) is expected to contain impurities that can make corrosion to structural metallic materials at elevated temperatures. According to the chemical thermodynamics and kinetics, the carbon activity of helium can be calculated, and it is indicated that a high “CH₄/H₂O” ratio may lead to severe carburizing of the alloys. On this basis, corrosion tests were conducted on the three heat-resistance alloys Inconel 617, Hastelloy X, and Incoloy 800H at 950°C using helium environment with impurities, and mainly the effect of carburization was investigated. The corrosion samples were observed by Scanning Electron Microscopy (SEM) with Energy Disperse Spectroscopy (EDS), Electron Probe Microanalyzer (EPMA), and Carbon-sulfur Analyzer. These all alloys showed the oxidation and carburizing behavior, in which the carburized depth of Hastelloy X was shallow due to the dense oxide scale.

Abstract: Nuclear energy is one of the most promising energy sources to satisfy energy security, environmental protection, and efficient supply. The High Temperature Gas-cooled Reactor (HTGR) has attractive inherent safety features and it can be used as many kinds of heat applications such as hydrogen production, electricity generation, process heat supply, district heating and desalination. Many countries, especially developing countries, show their interests in HTGR. Graphite materials are used for the core components of the HTGR. IG-110 graphite, fine-grained isotropic graphite, with high strength and high oxidation resistance is used in the High temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) and High Temperature Reactor-Pebble-bed Modules (HTR-PM) in China. IG-110 graphite is a major candidate for the core graphite components of the Very High Temperature Reactor (VHTR) which is one of HTGRs and one of the most promising candidates as the Generation-IV nuclear reactor systems. This paper describes design of core components of HTTR and R&D on nuclear graphite for HTGR. JAEA established the graphite structural design code and inspection standard of graphite to construct the HTTR. JAEA developed an in-service inspection method and a draft graphite structural design code for future HTGR on the basis of the HTTR technologies. Moreover, JAEA are now developing the design data base of IG-110 graphite and IG-430 graphite including irradiation data for HTGR.

Abstract: Carbon-14 is a radionuclide, which is a by-product in the operation of various nuclear reactor facilities. it also came from the interaction of cosmic ray with nitrogen and hydrogen in the atmosphere globally. This article elaborates the source of the ¹⁴C in High-Temperature Gas-Cooled Reactor, the amount of ¹⁴C released to the environment, as well as the forms of carbon-14. Meanwhile, the author presents the environment impact of this radionuclide. This paper concluded that ¹⁴N (n, p) ¹⁴C reaction in the fuel spheres and coolant gas is major source, and CO₂ is major release form. The conclusion could provide the references and suggestions for storage, disposal and release reduction of this type of waste.

Abstract: High temperature gas cooled reactor (HTGR) with higher outlet coolant temperature nearly 1000°C is expected for direct utilization of process heat to hydrogen production. The thermal analysis of reactor internals with 3 dimensional, flow paths coupled model was conducted to demonstrate how strictly PSR block gaps must be closed to limit core bypass flow rate ratio lest fuel temperature should exceed admissible level, and the highly heat resistant core restraint mechanism must be developed in consequence. Potential applicability of the core restraint mechanism made of C/C composite, the attractive candidate material, was demonstrated by point design with adequate thickness and FEM stress analysis for material with orthotropic anisotropy .

Abstract: Graphite materials are used for structural components in the core of high temperature gas-cooled reactors (HTGRs) because of their excellent thermo/mechanical properties. When the core temperature is raised at an accident, the thermal stress of the components is induced, and it enhances the fracture probability of them. In general, the thermal conductivity of graphite is decreased by neutron irradiation due to irradiation-induced defects preventing heat conduction by phonon. It is hence expected that decreased thermal conductivity is recovered to some extent by thermal annealing at the accident. Therefore, the consideration of the thermal annealing effect is placed as much important subject in the fracture/strength evaluation of the graphite components at the accident. In the present study, the thermal stress and the fracture probability of graphite components influenced by the thermal annealing were investigated by a finite element method (FEM) analysis. It was shown that the annealing effect decreases the thermal stress and a certain level of the fracture probability.

Abstract: The carbon fiber reinforced carbon-carbon composite (C/C composite) is one of the candidates due to its excellent thermal stability as well as high strength. A two-dimensional C/C composite has great anisotropy in those properties in with- and across- fiber directions. It is, therefore, important to consider the anisotropy for the stress evaluation and for the fracture probability of the components. In the present study, FEM analyses on deformation and stress of the component were carried out taking account of the anisotropy. In addition, the fracture probability of the components was evaluated by the statistical fracture theory. It was found that anisotropy affect the thermal stress and the risk of rupture.

Abstract: As an advanced in-core material in high temperature gas-cooled reactors (HTGRs), superplastic ceramics is attractive due to the possibility of the plastic working. For the application to the nuclear fields, the basic concept of design criteria was studied for typical superplastic ceramics, tetragonal zirconia polycrystals containing 3mol% yttria (3Y-TZP). The experimental results on 3Y-TZP showed that it is possible to apply the Weibull weakest-link theory to decide the stress limits in the criteria. The Weibull parameter m was evaluated as 9.5 for the bending and as 26.5 for the compressive. The applicability of the Weibull theory was also verified by the bending test results with different span. Based on the graphite structural design guidelines for the High Temperature Engineering Test Reactor (HTTR), the design stress limits for 3Y-TZP was proposed. It was shown that the proposed stress limits have appropriate safety margin and thought to be effective to evaluate the integrity of in-core structure made of 3Y-TZP.