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Journal Articles

Study on operation scenario of tritium production for a fusion reactor using a high temperature gas-cooled reactor

Kawamoto, Yasuko*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Katayama, Kazunari*; Goto, Minoru; Nakagawa, Shigeaki

Fusion Science and Technology, 68(2), p.397 - 401, 2015/09

 Times Cited Count:1 Percentile:12.03(Nuclear Science & Technology)

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. Herein, methods for supplying a fusion reactor with tritium are discussed. Use of a high temperature gas cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is periodically exchanged (batch) using the block type HTGR. In the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed type HTGR (PBMR) and the block type HTGR (GTHTR300) are assumed as the calculation and comparison targets. Simulation is made using the continuous-energy Monte Carlo transport code MVPBURN. It is shown that the continuous operation using the pebble bed type HTGR has almost the same tritium productivity compared with the batch operation using the block type HGTR. The issues for pebble bed type HTGR as a tritium production device are discussed.

Journal Articles

Development of simple method to incorporate out-of-core cooling effect on thorium conversion in multi-pass fueled reactor and investigation on characteristics of the effect

Fukaya, Yuji

Annals of Nuclear Energy, 81, p.301 - 305, 2015/07

 Times Cited Count:1 Percentile:12.03(Nuclear Science & Technology)

Development of a simple method to incorporate the out-of-core cooling effect on the thorium conversion in multi-pass fueled reactors and investigation on characteristics of the effect have been performed. For multi-pass fueled reactors, such as Molten Salt Breeder Reactor (MSBR) and Pebble-Bed Modular Reactor (PBMR), fuel moves in the core and exits from the core. The nuclides decay also out of the core, and it should be also considered if it is important for core characteristics. In the present study, $$^{233}$$Pa is considered to evaluate the thorium conversion accurately. To take the effect into account, in the present study, an effective decay constant is proposed to make equilibrium concentration of $$^{233}$$Pa without out-of-core cooling equal to that of out-of-core cooling. With the effective decay constant, the out-of-core cooling effect can be incorporated even with the code system using macroscopic cross sections generated by cell burn-up calculations without any code modification. In addition, the characteristic of out-of-core cooling effect for the thorium conversion is evaluated for thorium fueled reactors of MSBR and PBMR. It is concluded that the out-of-core cooling effect is suitable for MSBR to enhance thorium conversion because of the fast flow rate of fuel salt. On the other hand, the effect is not important and not realistic to employ for PBMR because the in-core residence time of approximately 100 days is longer than the half-life of $$^{233}$$Pa of 27.0 days, and the effect cannot improve the conversion ratio drastically.

Journal Articles

Effective thermal conductivity of a compressed Li$$_2$$TiO$$_3$$ pebble bed

Tanigawa, Hisashi; Hatano, Toshihisa; Enoeda, Mikio; Akiba, Masato

Fusion Engineering and Design, 75-79, p.801 - 805, 2005/11

 Times Cited Count:36 Percentile:91.6(Nuclear Science & Technology)

In order to elucidate thermo-mechanical behaviour of the blanket with the water-cooled ceramic breeders, the effective thermal conductivity of a compressed Li$$_2$$TiO$$_3$$ pebble bed was measured by the hot wire method. The measurement chamber for the thermal conductivity was inserted into a tensile test-apparatus. Under controlled temperature, atmosphere and mechanical loads, the stress-strain property and the effective thermal conductivity of the pebble beds were measured simultaneously. At the temperatures ranging from 673 to 973K, increases of the effective thermal conductivity due to the compressive deformation were confirmed. In addition, it was found that history of the mechanical and thermal loads on the bed affected the thermo-mechanical properties of the pebble bed.

JAEA Reports

Proceedings of the 11th International Workshop on Ceramic Breeder Blanket Interactions; December 15 - 17, 2003, Tokyo, Japan

Enoeda, Mikio

JAERI-Conf 2004-012, 237 Pages, 2004/07

JAERI-Conf-2004-012.pdf:44.1MB

This report is the Proceedings of "the Eleventh International Workshop on Ceramic Breeder Blanket Interactions" which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li$$_{2}$$TiO$$_{3}$$ and other breeders, fabrication technology developments and characterization of the Li$$_{2}$$TiO$$_{3}$$ and Li$$_{4}$$SiO$$_{4}$$ pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li$$_{2}$$TiO$$_{3}$$ and Li$$_{4}$$SiO$$_{4}$$ pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system.

Journal Articles

Fusion power reactor designs adopting SiC/SiC composite as the structural material

Nishio, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 80(1), p.14 - 17, 2004/01

SiC/SiC composite is a promising structural material candidate for fusion power cores and has been considered internationally in several power plant studies. It offers safety advantages arising from its low induced radioactivity and afterheat, and the possibility of high efficiency of energy conversion through high temperature operation. The latest SiC/SiC-based power core design studies are summarized, and the key SiC/SiC parameters affecting the performance of power core components are highlighted.

Journal Articles

Development of supercritical pressure water cooled solid breeder blanket in JAERI

Akiba, Masato; Ishitsuka, Etsuo; Enoeda, Mikio; Nishitani, Takeo; Konishi, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 79(9), p.929 - 934, 2003/09

no abstracts in English

JAEA Reports

Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

Kosaku, Yasuo; Kuroda, Toshimasa*; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Sato, Shinichi*; Osaki, Toshio*; Miki, Nobuharu*; Akiba, Masato

JAERI-Tech 2003-058, 69 Pages, 2003/06

JAERI-Tech-2003-058.pdf:5.86MB

The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket.

JAEA Reports

Thermal cycle test of elemental mockups of ITER breeding blanket

Yanagi, Yoshihiko*; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Enoeda, Mikio; Akiba, Masato

JAERI-Tech 2002-046, 45 Pages, 2002/05

JAERI-Tech-2002-046.pdf:2.61MB

no abstracts in English

Journal Articles

Safety analysis of ITER test blanket module for water cooled blanket with pebble bed breeder

Enoeda, Mikio; Kuroda, Toshimasa*; Moriyama, Koichi*; Ohara, Yoshihiro

Journal of Nuclear Science and Technology, 38(11), p.921 - 929, 2001/11

 Times Cited Count:2 Percentile:20.38(Nuclear Science & Technology)

Test module testing in ITER is one of the most important mile-stone for development of the DEMO blanket. In the design of test modules in ITER, it is very important to show that test modules do not cause additional safety concern to ITER. This work has been performed for the evaluation of the substantial safety of Test Module of Water Cooled Solid Blanket, which is the current candidate blanket for the DEMO blanket in Japan. Major issues of the evaluation were establishment of post accident cooling in TM, hydrogen gas generation by Be-steam reaction, and pressure increase and spilled water amount by Loss of Coolant Accident (LOCA) event. The evaluation was performed to derive the upper bound of consequences in significant events, of which scenario can be assumed by the similarity of the safety analysis of Shielding Blanket.

Journal Articles

Development of ceramic breeder blankets in Japan

Takatsu, Hideyuki; Kawamura, Hiroshi; Tanaka, Satoru*

Fusion Engineering and Design, 39-40, p.645 - 650, 1998/09

 Times Cited Count:17 Percentile:78.96(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Design development of breeding blanket based on pebble bed concept for fusion experimental reactor

Miura, H.*; *; *; Takatsu, Hideyuki; Kuroda, Toshimasa*; Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kurasawa, Toshimasa; *; et al.

Fusion Technology 1996, 0, p.1339 - 1342, 1997/00

no abstracts in English

Journal Articles

Reactor-grade plutonium burning by pebble bed type HTGRs

Fujimoto, Nozomu; Yamashita, Kiyonobu

Proc. of Int. Conf. on Future Nuclear Systems (Global'97), 2, p.957 - 962, 1997/00

no abstracts in English

JAEA Reports

Study on temperature coefficients of reactivity for pebble bed type HTGRs loaded with highly $$^{239}$$Pu-enriched plutonium

*; Yamashita, Kiyonobu; Shindo, Ryuichi; Fujimoto, Nozomu

JAERI-Tech 96-025, 50 Pages, 1996/06

JAERI-Tech-96-025.pdf:1.3MB

no abstracts in English

Journal Articles

HTGR type minor actinide transmutation reactor

Fujimoto, Nozomu; Yamashita, Kiyonobu; *

Nihon Genshiryoku Gakkai-Shi, 38(4), p.304 - 306, 1996/00

no abstracts in English

JAEA Reports

Convertible shielding to ceramic breeding blanket

Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; *; Hashimoto, T.*; Kuroda, Toshimasa*; Takatsu, Hideyuki

JAERI-Tech 95-031, 19 Pages, 1995/05

JAERI-Tech-95-031.pdf:0.72MB

no abstracts in English

Journal Articles

Analysis of control rod reactivity worths for AVR power plant at cold and hot conditions

Yamashita, Kiyonobu; Murata, Isao; Shindo, Ryuichi; *; H.Werner*

Journal of Nuclear Science and Technology, 31(5), p.470 - 478, 1994/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Layered pebble bed concept for ITER breeding blanket

Takatsu, Hideyuki; Mori, Seiji*; Yoshida, Hiroshi; Hashimoto, T.*; Kurasawa, Toshimasa; Koizumi, Koichi; Enoeda, Mikio; Sato, Satoshi; Kuroda, Toshimasa*; *; et al.

Fusion Technology 1992, p.1504 - 1508, 1993/00

no abstracts in English

JAEA Reports

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