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Fuel Behavior and Fission Product Release under HTGR Accident Conditions
Published in J. T. Rogers, Fission Product Transport Processes in Reactor Accidents, 2020
Kousaku Fukuda, Kimio Hayashi, Koreyuki Shiba
In early 1 989 a final decision was made over construction of a 30 MWth HTGR called as High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel having been developed at Japan Atomic Energy Research Institute, JAERI, is a “pin-in-block” type fuel element which is com posed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod.
Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs
Published in Nuclear Technology, 2022
Pavlo Ivanusa, Philip Jensen, Caitlin A. Condon, Amoret L. Bunn
Overall, three different reactor cores were used in the comparison: a helium-cooled prismatic core, a helium-cooled pebble bed reactor (PBR), and a fluoride-lithium-beryllium (FLIBE) molten-salt-cooled PBR. All three reactor designs provide their own benefits and are possible technologies. All of the reactors use as their uranium fuel. The PBR and FLIBE fuels embed the TRISO particles within a much larger pebble of graphite. The high-temperature engineering test reactor (HTTR) fuel uses small fuel compacts made of graphite, and the TRISO particles are then embedded in these fuel compacts. The fuel compacts are further stacked within fuel channels in a much larger graphite hexagonal block. Each one of these designs was then analyzed over a range of different fuel enrichments.
Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident
Published in Journal of Nuclear Science and Technology, 2018
Shuichiro Miwa, Yasunori Yamamoto, Go Chiba
Researches and developments on HTGR have been carried out for a number of years in Japan using high-temperature engineering test reactor (HTTR). HTTR in JAEA is the first HTGR in Japan with a thermal power of 30 MW and maximum coolant (helium) outlet temperature of 950 °C [209]. Not only as a power generation, due to its utilization of high-temperature coolant, HTGR can be potentially utilized for hydrogen production by coupling with a chemical plant using iodine–sulfur process (IS process), and a set of safety requirement to satisfy the regulations was presented by Sato et al. [210]. Various safety evaluation tests using HTTR have been conducted at JAEA including thermal-load fluctuation test, reactivity temperature coefficient measurement and loss of forced cooling (LOFC) test [211]. New advanced reactor named naturally safe high-temperature gas-cooled reactor (NSHTR) concept was also developed at JAEA after the Fukushima accident [212].