Network


Latest external collaboration on country level. Dive into details by clicking on the dots.

Hotspot


Dive into the research topics where Chunhe Tang is active.

Publication


Featured researches published by Chunhe Tang.


Nuclear Engineering and Design | 2002

Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor

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

Research and Development of Fuel Element for Chinese 10 MW High Temperature Gas-cooled Reactor

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

Preparation of UO2 Kernel for HTR-10 Fuel Element

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.


Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1 | 2008

Preliminary Results of the HFR-EU1 Fuel Irradiation of INET and AVR Pebbles in the HFR Petten

Alain Marmier; Michael A. Fütterer; M. Laurie; Chunhe Tang

The irradiation experiment HFR-EU1 in the HFR Petten is currently being conducted by the European Commission’s Joint Research Centre – Institute for Energy (JRC-IE). The irradiation targets are 5 spherical High Temperature Reactor (HTR) fuel pebbles, 2 of INET production and 3 of former German production. Both types are made of TRISO coated particles and are tested for their potential for very high temperature performance and high burn-up. The irradiation started on 29 September 2006 and, by 24 February 2008, had accumulated 12 reactor cycles totaling 332.8 efpd and a calculated maximum burn-up of 8.9% FIMA (INET) and 11.2% FIMA (AVR). The objective of the HFR-EU1 test is to irradiate 5 HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, improved fuel use) and thus reduced waste production. The surface temperature of all pebbles was held constant during the irradiation, with the exception of HFR downtime and power transients. HFR-EU1 should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: • high fuel surface temperature of 900°C (INET) and 950°C (AVR); • very high burn-up of 17% FIMA (INET) and 20% FIMA (AVR) which is significantly higher than the license limit of the HTR-Modul (approx. 8% FIMA); it will be explained in this paper why this objective had to be somewhat reduced due to excessive irradiation time requirements and technological difficulties. This paper provides the irradiation history of the experiment performed so far including data on fission gas release.© 2008 ASME


Science and Technology of Nuclear Installations | 2018

Study on the Comprehensive Properties and Microstructures of A3-3 Matrix Graphite Related to the High Temperature Purification Treatment

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

Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength

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.


Science and Technology of Nuclear Installations | 2017

The Electric Current Effect on Electrochemical Deconsolidation of Spherical Fuel Elements

Xiaotong Chen; Zhenming Lu; Hongsheng Zhao; Bing Liu; Junguo Zhu; Chunhe Tang

For High-Temperature Gas-Cooled Reactor in China, fuel particles are bonded into spherical fuel elements by a carbonaceous matrix. For the study of fuel failure mechanism from individual fuel particles, an electrochemical deconsolidation apparatus was developed in this study to separate the particles from the carbonaceous matrix by disintegrating the matrix into fine graphite powder. The deconsolidated graphite powder and free particles were characterized by elemental analysis, X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectrometer (EDS), and ceramography. The results showed that the morphology, size distribution, and element content of deconsolidated graphite matrix and free particles were notably affected by electric current intensity. The electrochemical deconsolidation mechanism of spherical fuel element was also discussed.


Journal of Nuclear Science and Technology | 2017

Modification in the stress calculation of PyC material properties in TRISO fuel particles under irradiation

Rong Li; Bing Liu; Chunhe Tang

ABSTRACT Tristructural-isotropic (TRISO) coated fuel particles are a key component in the high-temperature gas-cooled reactor fuel elements. The stresses in a TRISO-coated particle depend upon the coating layer properties which are affected by neutron irradiation. The modifications of pyrolytic carbon (PyC) material properties under irradiation require the use of time-dependent material parameters in the stress calculations, which was performed with an analytic solution in this study. The experimental results indicated that PyC density increased at the early stages of irradiation while PyC anisotropy increased obviously at the later stages of irradiation, both of which caused particle stresses to increase. The PyC creep coefficient was modeled as increasing with neutron fluence. This significantly decreased the silicon carbide and inner pyrolytic carbon tangential and interface radial stresses. Increasing the Youngs modulus of the PyC had little effect on the particle stresses. As a result, compared with constant material properties, the dynamic variation of PyC material parameters under irradiation further improved particle stress calculations and made the stress model closer to actual reactor conditions.


International Journal of Minerals Metallurgy and Materials | 2009

Purity of SiC powders fabricated by coat-mix

Limin Shi; Hongsheng Zhao; Chunhe Tang

Abstract Silicon carbide powders were synthesized by the coat-mix process, with phenolic resin and silicon powders as starting materials. The effects of synthetic conditions, including sintering temperature and the molar ratio of resin-derived carbon to silicon on the composition and the purity of the resultant powders were investigated. The results show that a higher sintering temperature and an appropriate molar ratio of resin-derived carbon to silicon are favorable for producing high purity silicon carbide powders. It is found that the silicon carbide content increases slightly with increasing the sintering temperature during the solid-solid reaction. The temperature gradient plays an important role on this trend. When the sintering temperature is raised up to 1500°C, the formation of silicon carbide is based on the liquid-solid reaction, and high purity (99.8wt%) silicon carbide powders can easily be obtained. It can also be found that the optimum molar ratio of resin-derived carbon to silicon is 1:1.


Journal of Nuclear Materials | 2014

High temperature oxidation behavior of SiC coating in TRISO coated particles

Rongzheng Liu; Bing Liu; Kaihong Zhang; Malin Liu; Youlin Shao; Chunhe Tang

Collaboration


Dive into the Chunhe Tang's collaboration.

Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Top Co-Authors

Avatar
Researchain Logo
Decentralizing Knowledge