Preliminary Test Results for Post Irradiation Examination on the HTTR Fuel

The Japan Atomic Energy Agency (JAEA) has carried out research and development of high-temperature gas-cooled reactors (HTGRs) since 1960s. The high-temperature engineering test reactor (HTTR) has attained full power in 2000. In this study, a failure probability calculation method for Japanese highquality HTGR fuel particles is suggested for the design of Japanese HTGRs. Failure probability is strongly dependent on the irradiation characteristics of pyrolytic carbon (PyC) layers of HTGR fuel particles. However, it is very di‹cult to obtain new irradiation data of PyC layers with experiments. Therefore, in this study, we calculated failure probabilities using a report summarizing the irradiation data obtained at various irradiation temperatures of PyC layers with various as-fabricated characterisitics and the result of the failure probability measurement of the irradiation test of HTTR fuel. The characteristics of the coating layers of Japanese HTGR fuels, including future fuel types, would be very similar to those of HTTR fuel. In addition, some calculations were carried out with the suggested method. It was shown that the thickening of the SiC layer is more eŠective in reducing the failure probability than the thickening of the buŠer layer. Thus, the thickening of the SiC layer is important for the modiˆcation of HTGR fuel for burnup extension, suppressing the increase in core volume.

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Development of Technology of Electrochemical Separation of Fuel Element of HTGR

Dyussambayev,

Gizatulin,

Shaimerdenov

et al. 2019

jour

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High-temperature gas-cooled reactor (HTGR) by two-thirds consists of graphite. Graphite is a moderator and reflector of neutrons, as well as the main structural material of the active zone, which allows operating the reactor at high temperatures Graphite is moderator and neutron reflector, and the main structural material of the core, allowing the reactor to operate at high temperatures. Nuclear fuel is a spherical microfuel, which are compressed into a graphite compact, from which fuel assemblies are then assembled. The Institute of Nuclear Physics of the Republic of Kazakhstan is working on the qualification of irradiated fuel HTGR. To study the properties of the irradiated fuel HTGR, it is necessary to extract microfuel from the graphite matrix. The paper presents the results of experiments on the electrochemical separation of graphite and unirradiated fuel HTGR.

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Preliminary Test Results for Post Irradiation Examination on the HTTR Fuel

Cited by 3 publications

References 5 publications

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

Suggested Failure Probability Calculation Method for Japanese High-Quality High-Temperature Gas-Cooled Reactor (HTGR) Fuel

Development of Technology of Electrochemical Separation of Fuel Element of HTGR

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