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Negyesi, M.; Amaya, Masaki
Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10
Times Cited Count:8 Percentile:57.23(Nuclear Science & Technology)Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi
JAEA-Technology 2014-038, 51 Pages, 2014/12
The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.
Fujii, Kimio
JAERI-Tech 2005-048, 108 Pages, 2005/09
The graphite-moderated power reactor was shut down in 1998 and its decommissioning program is being planned. Various graphites are used in the core of magnox-type reactors and HTTR as core-support structural materials and moderating materials of fast neutrons. For the nuclear graphite disposal, it is necessary to determine especially the treatment of long-lived nuclides, such as C which are generated in the graphite components during reactor operation. As a research, which solves the problem of the
C concentration, the cooperative research is concluded between JAERI and Japan Nuclear Power Corp. in 1999, and the research for the basic data acquisition has been advanced up to the present. To find the optimum conditions for
C reduction, basic data on oxidation reaction and the structure of graphite materials are indispensable. In the present experiment, we measure the air oxidation characteristics in the temperature range 450
800
C in Quality1 graphite and IG-110 graphite. Changes in pore diameter and pore size distribution due to air oxidation are discussed.
Sogabe, Toshiaki; Ishihara, Masahiro; Baba, Shinichi; Tachibana, Yukio; Yamaji, Masatoshi*; Iyoku, Tatsuo; Hoshiya, Taiji*
Materials Science Research International, 9(3), p.235 - 241, 2003/09
2D-C/C composite is one of the promising materials as a next-generation core material in gas-cooled reactors. Effect of air-oxidation on the thermal diffusivity up to 1673K of the 2D-C/C composite was investigated. The C/C composite specimens for measurement of thermal diffusivity were oxidized from 1 to 11 percent weight loss in air at 823K. Thermal diffusivity at room temperature declined 1020 percent for parallel to lamina direction and 5
9 percent for that of perpendicular within 11 percent weight loss by oxidation. Thermal diffusivity tended to decrease gradually as the increase of oxidation loss in parallel to lamina, however, it decreased in the beginning of oxidation pretty much and not so changed by further oxidation loss in perpendicular to lamina. Change in thermal conductivity under oxidation condition was also estimated.
; Hayashi, Kimio; Fukuda, Kosaku
JAERI-M 92-114, 20 Pages, 1992/08
no abstracts in English
; Ohashi, Kazutaka*; Iyoku, Tatsuo
FAPIG, 0(129), p.13 - 21, 1991/11
no abstracts in English
;
Journal of Nuclear Materials, 140, p.32 - 43, 1986/00
Times Cited Count:14 Percentile:80.31(Materials Science, Multidisciplinary)no abstracts in English
;
Journal of Nuclear Materials, 36(1), p.116 - 119, 1970/00
no abstracts in English
; Honda, Toshio*
Journal of Nuclear Science and Technology, 5(11), p.600 - 602, 1968/00
no abstracts in English