[en] Graphite is adopted as the major structural material in HTGR so the study of mechanical behaviour of graphite under high temperature, high radiation and oxidization becomes a main point. Physical and mechanical properties of graphite, fatigue behaviour of graphite and stress classification and failure criteria under the surroundings in question are discussed. The paper also gives an introduction to creeping and cracking problems of concrete in the stress analysis of HTGR pressure vessels (PCPV)

[en] This article describes the structural design requirements, structural arrangement and structural features of the ceramic and metallic internals of the 10 MW high-temperature gas-cooled reactor-test module (HTR-10). The graphite properties used in the ceramic internals are provided, along with the results of an operating stress analysis of the graphite components and the metallic components. Satisfactory results were obtained for the machining and installation of the ceramic components and the stress analysis of the graphite and metallic components of HTR-10

[en] The mechanical behavior of irradiated graphite fuel element is very effective on the safety and reliability of HTR operation. The analytical stress solution for graphite fuel element is given in this paper in the spherical coordinates on basis of the linear viscoelastic theory. The stress calculation of the fuel element under fast neutron irradiation is conducted during it flows through the HTR core. (author)

[en] A probability finite element assessment program was developed to evaluate he security of graphite components in the HTR-10 (10 MW high temperature as-cooled reactor-test module), based on the MARC non-linear finite element code and the strength uncertainty of the graphite material. Using user-defined subroutines (UDS), the irradiation thermal analysis subroutine, irradiation static analysis subroutine and probability assessment subroutine are embedded into the MARC program. The recompiled MARC program take into account irradiation-induced changes in graphite components such as the thermal conductivity coefficient, the thermal expansion coefficient, the creep coefficient, the elastic modulus, and the strength. The failure probabilities of the graphite components in the HTR-10, either under normal operating conditions or cold shutdown conditions, were evaluated. Additional analyses were done with the irradiation deformation increasing 20% and the creep coefficient decreasing 20%, to see the influence of irradiation deformation and the creep effect on the failure probability. The study showed that the probability finite element assessment method is an effective tool to assess the probability of structure failure

[en] Design principle, feature, structure, analysis and evaluation of fuel assembly for 5 MW THR are given

[en] The world's first nuclear reactor dedicated to providing district heating started operating in China in 1989. The NHR-5 is a JMW light water reactor cooled by natural circulation. It is designed to operate as a single phase reactor, that is a PWR, and with slight boiling. The reactor core is held in a hanging barrel containing 12 large fuel assemblies and 4 small fuel assemblies. The mechanical details of the assemblies and fuel rods are given. Since start-up, the NHR-5 has been successfully used in applications such as heat supply, refrigeration and desalination and is seen as an economic and clean heat source. (3 figures, 2 tables) (UK)

[en] The Strength of IG11 graphite used in the HTR-10 structure was measured experimentally. The strength distribution follows a normal distribution, which was used to estimate the graphite failure strength. The ultimate strength of the graphite was estimated from the results and the Weibull distribution was used to estimate the probability of the graphite failure

[en] The irradiation induced deformation and creep usually happens to graphite components irradiated by fast neutron influence in nuclear reactors, especially in the high temperature-gas-cooled reactors. Therefore, it is necessary and important to analyze and assess the mechanical behavior of graphite components under irradiation conditions during their design process. The irradiation induced deformation can be dealt with in the same way always used to deal with the thermal expansion but the irradiation induced creep should be analyzed by means of the viscoelastic model which is commonly analyzed by two methods, one of which is called the state equation method or the incremental method, and the other is the hereditary method. A FEM method using the hereditary method to analyze behavior of the irradiated graphite components is described in detail in this paper. It has the advantages of the hereditary method, and in the creep computation, the component volume is considered incompressible. The stress and deformation calculation of the graphite block of side reflector of HTR-10, high temperature reactor- test module in the Institute of Nuclear and New Energy Technology of Tsinghua University, Beijing, has been conducted, considering irradiation induced deformation and creep. (authors)

[en] The structure configuration and safety features for the Graphite Internals of 10 MW High-temperature Gas-Cooled Test Reactor are briefly described and the calculation results of stress and deformation of a typical side reflector graphite block under condition of fast neutron irradiation are given

[en] In 10 MW High Temperature Gas-cooled Reactor (HTR-10), the wear of graphite components generates the graphite dust to influence the normal working of the reactor. The graphite dusts come from three parts: core, unload pipe and load pipe. Based on the graphite wear experiments, authors here estimated that the graphite generation under normal working condition is about 2.74 kg/a. By the mass-weighted average, the distribution function of volume, area and diameter for graphite particle were obtained. (author)