Yaping Tang
Tsinghua University
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Featured researches published by Yaping Tang.
Nuclear Engineering and Design | 2002
Chunhe Tang; Yaping Tang; Junguo Zhu; Yanwen Zou; Jihong Li; Xiaojun Ni
Abstract The Chinese 10 MW high temperature gas-cooled reactor (HTR-10) attained its first criticality on December 21, 2000. The fabrication of the first fuel for the HTR-10 started in February 2000 at the Institute of Nuclear Energy Technology (INET), Tsinghua University. Up to September 2000, a total of 11u2008721 spherical fuel elements were successfully produced. The average free uranium fraction of the first fuel-determined by the burn-leach method-was 5.0×10 −5 . So far, the release rate R/B of the fission gas, measured in the irradiation test, shows that not a single particle in three irradiated spherical fuel elements failed as the results of the irradiation test carried out in Russia. This paper describes the design parameter, the fabrication technology and the performance data of the HTR-10 first fuel, and the production and quality control experiences obtained from the manufacture of the first fuel for the HTR-10.
Journal of Nuclear Science and Technology | 2000
Chunhe Tang; Yaping Tang; Junguo Zhu; Xueliang Qiu; Jihong Li; Shijiang Xu
The fuel elements for Chinese 10 MW High Temperature Gas-cooled Reactor (HTR-10) are spherical all- ceramic fuel elements (SFE). TRISO (Tri-isotropic) coated fuel particles (CP) are uniformly dispersed in the graphite matrix of the fuel element. All radiological fission products are almost completely retained inside the SiC layer of the intact CP. The fabrication of SFE includes UO2 kernel preparation by sol-gel method, pyrolytic carbon (PyC) and SiC coating on the UO2 kernels by Chemical Vapor Deposition, manufacture of SFE by the quasi-isostatic pressing and the inspection of over 30 kinds of properties. This paper describes research and development (R & D) and design specification of the HTR-10 SEF, summarizes the fabrication technology and quality control mastered in R & D for HTR-10 SFE.
Journal of Nuclear Science and Technology | 2004
Xiaoming Fu; Tongxiang Liang; Yaping Tang; Zhichang Xu; Chunhe Tang
A 10 MW high temperature gas-cooled reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, China. HTR-10 reached its first criticality successfully in December, 2000 and realized its full power operation in the beginning of 2003. Fabrication technology for HTR-10 spherical fuel element has been established and developed in INET through a lot of R&D activities during the past 20 years. A process known as total gelation process of uranium (TGU) has been developed to prepare UO2 kernel for HTR-10 fuel element in INET. The TGU process combined advantages of both external gelation and internal gelation method of conventional sol-gel process. The sol prescription consists of not only urea, polyvinylalcohol (PVA), tetrahydrofurfuyl alcohol (THFA), but also hexamethylene tetra-amine (HMTA) that is always employed in internal gelation process. The TGU process and facilities are described in this paper. Fuel kernel production for HTR-10 fuel element using this TGU process has completed up to April, 2002. This paper also reported the quality inspection data of kernel production.
New Carbon Materials | 2017
Xiangwen Zhou; Yaping Tang; Zhenming Lu; Jie Zhang; Bing Liu
Abstract Since its first successful use in the CP-1 nuclear reactor in 1942, nuclear graphite has played an important role in nuclear reactors especially the high temperature gas-cooled type (HTGRs) owing to its outstanding comprehensive nuclear properties. As the most promising candidate for generation IV reactors, HTGRs have two main designs, the pebble bed reactor and the prismatic reactor. In both designs, the graphite acts as the moderator, fuel matrix, and a major core structural component. However, the mechanical and thermal properties of graphite are generally reduced by the high fluences of neutron irradiation of during reactor operation, making graphite more susceptible to failure after a significant neutron dose. Since the starting raw materials such as the cokes and the subsequent forming method play a critical role in determining the structure and corresponding properties and performance of graphite under irradiation, the judicious selection of high-purity raw materials, forming method, graphitization temperature and any halogen purification are required to obtain the desired properties such as the purity and isotropy. The microstructural and corresponding dimensional changes under irradiation are the underlying mechanism for the changes of most thermal and mechanical properties of graphite, and irradiation temperature and neutron fluence play key roles in determining the microstructural and property changes of the graphite. In this paper, the basic requirements of nuclear graphite as a moderator for HTGRs and its manufacturing process are presented. In addition, changes in the mechanical and thermal properties of graphite at different temperatures and under different neutron fluences are elaborated. Furthermore, the current status of nuclear graphite development in China and abroad is discussed, and long-term problems regarding nuclear graphite such as the sustainable and stable supply of cokes as well as the recycling of used material are discussed. This paper is intended to act as a reference for graphite providers who are interested in developing nuclear graphite for potential applications in future commercial Chinese HTGRs.
New Carbon Materials | 2016
Xiangwen Zhou; Zhenming Lu; Xin-nan Li; Jie Zhang; Bing Liu; Yaping Tang
Abstract The effects of temperature on the oxidation behavior of the A3-3 matrix graphite (MG) in the temperature range 798-973 K in air with a flow rate of 100 ml/min to burn-offs of 10-15 wt%, were investigated by a home-made thermo-gravimetric experimental setup. The oxidation rate (OR) increases significantly with the temperature. The OR at 973 K is over 70 times faster than at 798 K. The oxidation kinetics of A3-3 MG in air at temperatures up to 973 K is in the reaction control regime, where the activation energy is 176 kJ/mol and the Arrhenius equation could be described as: OR =2.9673×10 8 ·exp(-21124.8/ T ) wt%/min. The relatively lower activation energy of MG than that of structural nuclear graphite indicates that MG is more easily oxidized.
Science and Technology of Nuclear Installations | 2018
Xiangwen Zhou; Zhenming Lu; Jie Zhang; Jing Song; Bing Liu; Yaping Tang; Chunhe Tang
At the beginning, a comparative analysis was made on the oxidation corrosion rate and ash content of A3-3 matrix graphite (MG) pebbles lathed before and after high temperature purification (HTP) treatment. Their oxidation corrosion rate and ash contents were almost identical, which indicated that the HTP process was to purify the entire MG pebbles and not limited on the surfaces. Furthermore, the multiple mechanical and thermal properties of MG treated without and with the treatment of HTP at ~1900°C were compared and their microstructure features were characterized as well. As the crush strength, oxidation corrosion rate, and erosion rate of MG without HTP treatment did not satisfy the specifications, the comprehensive properties and purity of MG with HTP were improved in various degrees through the HTP process so that all performances met the requirements of the A3-3 MG. The improvement of crush strength and erosion rate of MG in the HTP process could be mainly attributed to the upgradation of ordered microstructure and corresponding increase of density. However, the enhancement of oxidation corrosion rate was due to the synergistic effects of microstructural optimization and reduction of impurity elements, especially the transition metal elements of MG in the HTP process.
Science and Technology of Nuclear Installations | 2017
Xiangwen Zhou; Cristian I. Contescu; Xi Zhao; Zhenming Lu; Jie Zhang; Yutai Katoh; Yanli Wang; Bing Liu; Yaping Tang; Chunhe Tang
Matrix graphite (MG) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 MG at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized MG specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of MG. The transition temperature between Regimes I and II is ~700°C and the activation energy ( ) in Regime I is around 185u2009kJ/mol, a little lower than that of nuclear graphite, which indicates MG is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of MG than oxidation at 900°C. Comparing with the strength of pristine MG specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. Microstructure images of SEM and porosity measurement by Mercury Porosimetry indicate that the significant compressive strength loss of MG oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.
Journal of Luminescence | 2017
Xiaotong Chen; Lu Peng; Mengli Feng; Yu Xiang; Aijun Tong; Linfeng He; Bing Liu; Yaping Tang
Nuclear Engineering and Design | 2014
Xiangwen Zhou; Jie Zhang; Zhenming Lu; Yanwen Zou; Yaping Tang
Archive | 2012
Ying Quan; Yaping Tang; Bing Liu; Xiaotong Chen; Chunhe Tang; Xuanna Fu