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Journal Articles

CT injection experiment in JFT-2M

Ogawa, Hiroaki; Ogawa, Toshihide; Tsuzuki, Kazuhiro; Kawashima, Hisato; Kasai, Satoshi*; Kashiwa, Yoshitoshi; Hasegawa, Koichi; Suzuki, Sadaaki; Shibata, Takatoshi; Miura, Yukitoshi; et al.

Fusion Science and Technology, 49(2), p.209 - 224, 2006/02

 Times Cited Count:3 Percentile:25.05(Nuclear Science & Technology)

no abstracts in English

Journal Articles

High performance tokamak experiments with a ferritic steel wall on JFT-2M

Tsuzuki, Kazuhiro; Kimura, Haruyuki; Kawashima, Hisato; Sato, Masayasu; Kamiya, Kensaku; Shinohara, Koji; Ogawa, Hiroaki; Hoshino, Katsumichi; Bakhtiari, M.; Kasai, Satoshi; et al.

Nuclear Fusion, 43(10), p.1288 - 1293, 2003/10

 Times Cited Count:39 Percentile:75.57(Physics, Fluids & Plasmas)

no abstracts in English

JAEA Reports

The Report of inspection and repair technology of sodium cooled reactors

Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru; Kasai, Shigeo; Soman, Yoshindo; Shimakawa, Yoshio; Hori, Toru; Chikazawa, Yoshitaka; Miyahara, Shinya; Hamada, Hirotsugu; et al.

JNC TN9400 2003-002, 109 Pages, 2002/12

JNC-TN9400-2003-002.pdf:8.12MB

Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. (1)Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow down system must be evaluated to realize its performance. (2)In-service inspection (ISI&R). The viewpoint of the commercialized plant's ISI&R was organized by comparing with the prototype reactor's ISI&R method. We also investigated short-term ISI&R methods without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize them with the present technology. Hereafter, the ISI&R of the commercialized plants must be defined by considering its characteristics. (3)Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was ...

Journal Articles

Feasibility Studies on Commercialized Fast Breeder Reactor System(1)-Sodium Cooled Fast Breeder Reactor

Nibe, Nobuaki; Shimakawa, Yoshio; Kasai, Shigeo;

Final Program and Ab, 0 Pages, 2001/00

None

JAEA Reports

JFY 1995 Progress report of conceptual design study on the recycle reactor; Study of core design

Naganuma, Masayuki; Kasai, Shigeo; Hayashi, Hideyuki; Mukaibou, Ryuichi

PNC TN9430 96-006, 157 Pages, 1996/07

PNC-TN9430-96-006.pdf:9.02MB

As the goal requested to FBR in the future practical time, there is the improvement of economy and safety, moreover, the decrease of risk to environment by burning the long-lived nuclides and the rise of nuclear non proliferation by preventing accumulation of Pu by dealing flexibly with supply and demand of Pu will be requested. To attain these goals, it is requested that the core has the capability to receive flexibly Actinide (Minor Actinide and Pu). In other words, the concept of the core (Recycle Core) that corresponds to recycle technology (Advanced Recycle Technology), receives various fuels (including MA, low decontaminated, high Pu enriched and so on) and can permit change of nuclear property and safety characteristic by receiving such fuels is demanded. In 1995' work, to examine the concept ofthis new core, authors conducted design studies about the following contents, and evaluated characteristics of these cores. (1)Design study about MOX ductless fueled core. In adopting the ductless fuel concept, the nuclear property is improved by the increase of volume ratio of fuel, and amount of the solid waste is decreased by the deletion of wrapper tube. Thus, the economic feature is improved and the risk to environment is decreased. In such reason, we designed MOX ductless fueled core (power : 600MWe, the cycle length : 18 months, Average burnup : 150000MWd/t) as the reference core of Recycle Reactor, and estimated some performances of this core. (1)In the ductless fueled core, the volume ratio of fuel became 13% larger than the duct fueled core, and that effect made breeding ratio increase 0.08 and burnup reactivity improve about 40%. (2)By adding 3% of MA (with 20% RE of MA), Pu enrichment and breeding ratio hardly change and burnup reactivity was improved about 30%. But Doppler coefficient and Na void reactivity became about 20% worse, and this reactor got severer in respect of safety. (3)ATWS analysis was conducted to the reference core, and ...

JAEA Reports

JFY 1995 Progress report of the development on the actinide recycle test reactor(ARTR)

Kasai, Shigeo; Tozawa, Katsuhiro; Akatsu, Minoru; Ogawa, Shinta; Watanabe, Ichiro; Hayafune, Hiroki; Naganuma, Masayuki; Ichimiya, Masakazu; Hayashi, Hideyuki; Mukaibou, Ryuichi

PNC TN9430 96-004, 152 Pages, 1996/07

PNC-TN9430-96-004.pdf:6.15MB

Authors are studying the Actinide Recycle Fast Breeder Reactor (named ARFBR in this paper), which contribute to the reduction of burdens to environments and to enhance the capability to prevent the nuclear proliferation as the entire nuclear recycle system (named Advanced Fuel Recycling FBR system (AFRFS) in this paper), and also investigating the ARTR for developing the ARFBR. The investigation of the ARTR consists of the design study of the ARTR and R&Ds of key technology existing in ARTR concept. The conceptual design study of the ARTR is planed to be conducted for 2 years from 1995 to 1996 as first stage. 1995's design study have been performed with drawing over all plant concept with supposing various tests in reactor and usage of reactor. Followings are distinctive feature of 1995's design study. (1)Maximum reactor power is 400MWt with about 1.6m diameter irradiation (burning) cores, which are designed to be operated up to 150GWd/t as average burn up. Maximum core diameter is about 2.5m for low power nuclear physics tests which are designed to be able to estimate characteristics of large scale core by using the test results. (2)Mixed oxide (MOX) and Mixed nitride (MN) core is designed respectively to be able to be used for static nuclear physics test, for nuclear and thermal transient test, and for full power irradiation or burning test. Each core is designed to terminate ATWS events passively, with using GEM for MOX core and with using spectral adjustment for MN core. (3)Fuel assembly is employed ductless type which is a promising candidate for the ARFBR. Sizing of a fuel assembly is determined in basis on MOX fuel design because MOX fuel pin length covers MN fuel pin which accommodates lesser FP gases because of its lower temperature. Fuel assembly is managed to be held by hydraulic force in case of freeing mechanical stopper by requirement of testability. (4)Reactor assembly is designed based on so called Head Access Loop Type Reactor. Main changes ...

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