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Journal Articles

Distance information display system for supporting decommissioning work of nuclear power plants

Miki, Naoya*; Harazono, Yuki*; Ishii, Hirotake*; Shimoda, Hiroshi*; Koda, Yuya

International Electronic Journal of Nuclear Safety and Simulation (Internet), 9(2), p.162 - 171, 2018/12

Kyoto University reports on the distance information display system developed in collaboration between "Fugen" and Kyoto University. In the distance information display system, the worker can easily measure the distance of the object. In this presentation, we report the test results carried out with "Fugen" and the questionnaire results of the subjects.

Journal Articles

Experiences on research reactors decommissioning in the NSRI of the JAEA

Tachibana, Mitsuo; Kishimoto, Katsumi; Shiraishi, Kunio

International Nuclear Safety Journal (Internet), 3(4), p.16 - 24, 2014/11

Three research reactors were permanently shut down in the Nuclear Science Research Institute (NSRI) of the Japan Atomic Energy Agency (JAEA) as of October 2014. Safe storage or one-piece removal method was applied to decommissioning of these research reactors depending on decommissioning cost and utilization of facilities and so on. Various kinds of data and experiences were obtained through decommissioning of these research reactors. This report shows data and experiences on the research reactors decommissioning in the NSRI of the JAEA.

Journal Articles

Fundamental experiment on the distance for fragmentation of molten core material during core disruptive accidents in sodium-cooled fast reactors

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

International Electronic Journal of Nuclear Safety and Simulation (Internet), 4(4), p.272 - 277, 2013/12

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents (CDAs) in sodium cooled fast reactors, fundamental experiments were conducted using a high-density melt and water as simulants for the molten fuel and coolant, respectively. The melt was discharged into a water pool through a nozzle (inner diameter: from 30 mm to 150 mm) under a simulated CDA condition where a liquid-liquid direct contact is maintained between the melt and water. The present results showed that measured distances for fragmentation were limited to approximately 10 percent of predictions by the existing representative correlation, and that vapor expansion with pressure buildup near the melt could facilitate the fragmentation and thus contribute to the reduction of the distance for fragmentation. Through the fundamental experiments, useful knowledge was obtained for the future development of an evaluation method.

Journal Articles

Plant maintenance and dismantling work support based on three-dimensional scanning technology

Ishii, Hirotake*

International Electronic Journal of Nuclear Safety and Simulation (Internet), 4(2), p.97 - 104, 2013/06

Recent progress of computer vision and micro-fabrication technologies has greatly reduced the cost and difficulty of three-dimensional modeling of maintenance and decommissioning work environments. This article presents recent progress of three-dimensional scanning technologies and their possible application to enhance safety and efficiency during plant maintenance and decommissioning.

Journal Articles

Failure and its mechanism of LWR and research reactor fuels

Yanagisawa, Kazuaki

International Electronic Journal of Nuclear Safety and Simulation (Internet), 1(3), p.246 - 257, 2010/09

A LWR fuel consisted of UO$$_{2}$$ pellet with zircaloy cladding and a research reactor fuel consisted of aluminide or silicide fuel cores with Al alloy. (1) LWR fuels were irradiated at the HBWR up to burn-up of 20 MWd/kgU and power ramped to have the PCI failure. The failure threshold of LWR was lower than that of HBWR. The in-core data was useful for obtaining the input data for the computer code FEMAXI. (2) Fresh or pre-irradiated LWR fuels were pulse-irradiated at the NSRR. For the fresh fuel, a failure occurred at 260 cal/g with cladding melt-brittle mechanism. For the pre-irradiated PWR fuel, a cladding split along one generatrix occurred at 118 cal/g. The failure mechanism may be the strong PCMI combined with the transient FGR. (3) Research reactor fuels failed by the through-plate cracking (les than 640 $$^{circ}$$C) or cladding melt (above 640 $$^{circ}$$C). A failure mechanism for the former was a tensile stress caused by a local uneven temperature profile during a quench.

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