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JAEA Reports

Design study on a fuel handling system in a sodium cooled reactor; Study in FY2004 (Joint research)

Chikazawa, Yoshitaka; Usui, Shinichi; Konomura, Mamoru; Ikeda, Hirotsugu

JAEA-Research 2006-032, 202 Pages, 2006/04

JAEA-Research-2006-032.pdf:38.12MB

In the feasibility study on commercialized fast breeder cycle system, fuel handling systems for sodium cooled reactors has been studied. In FY 2004 study, a fuel handling system with an EVST for a twin large scale reactor power plant is designed and key issues about the system are identified. A manipulator type fuel handling machine suitable for the upper internal structure with a slit designed and seismic analyses show that it can treat spent fuels without interaction with upper internal structure in earthquakes. Fuel handling time is reduced adopting a sodium pot which can carry 2 subassemblies in onetime. Spent fuels are stored at an EVST while their decay heat are reduced to be 5kW/subassembly. A new fuel handling system for fuels with minor actinide is designed considering 1kW/subassembly heat and shielding. A innovative concept without an EVST is also studied. A fuel handling system adopting fuel transfer without a sodium pot is constructed to reduce material mass. A fuel handling system for a metal fuel reactor plant has been design. From the result of a survey on a gas storage, a water pool storage with helium cans and EVST, a system with EVST is selected because of its economical and safety advantage. Fuel handling condition is briefly reviewed considering commercialized reactor fuel specifications such as minor actinide content and ODS cladding.

JAEA Reports

Design study on sodium-cooled reactor; Results of the studies in 2004 (Joint research)

Hishida, Masahiko; Murakami, Tsutomu*; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato*; Hayafune, Hiroki; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu; Uno, Osamu; et al.

JAEA-Research 2006-006, 125 Pages, 2006/03

JAEA-Research-2006-006.pdf:11.55MB

In Phase I of the "Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase II, design improvement for further cost reduction and the establishment of the plant concept has been performed. In this study, reactor core design and large-scale plant design have been performed by adopting the modified fuel assembly with inner duct structure and double-wall straight tube steam generator (SG), which concepts were chosen at the interim review of FY 2003. For this SG, safety logics have been studied and the structural concept has been established. And the plant designs improving the in-service inspection (ISI) and repair capability have been performed. Furthermore, elaborate confirmation of the design has been performed reflecting the development of elemental technology, back-up concepts have been proposed. Besides, cost reduction measures have been studied by reducing reactor grade materials, introducing autonomous standardizations, simplifying the design due to deregulation and adopting systemized standards for BOP and NSSS. From now on, reflecting the results of elemental experiments, in-depth design studies and examination of critical issues will be carried out and the plant concept will accomplish in preparation for the final evaluation in Phase II.

Journal Articles

Conceptual design study of helium cooled fast reactor in the "feasibility study" in Japan

Okano, Yasushi; Naganuma, Masayuki; Ikeda, Hirotsugu; Mizuno, Tomoyasu; Konomura, Mamoru

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

Conceptual design of gas-cooled fast reactor (GFR) have been studied for selecting applicable combinations of coolant, fuel material and configuration, and, balance of plant, as a part of feasibility study in Japan. Large-scale He-cooled GFR, employing mixed nitride fuel and achieving a high core outlet temperature of 850$$^{circ}$$C, is recognized to achieve attractive features as a future nuclear reactor system. Three fuel configurations are considered and compared in their core and safety performances; one is horizontal-flow cooling fuel assembly (F/A), another is hexagonal matrix block F/A, and the last one is sealed pin bundle F/A. The horizontal-flow and matrix block F/A cores show nearly the same neutronics performances on discharge burnup around 120GWd/ton, breeding ratio above 1.1, and, core cooling performances under depressurization condition without control rod scram or auxiliary core cooling system (ACCS) actuations; whereas around 30% smaller quantity of fissile Pu required is a merit for matrix concept. The sealed pin bundle F/A core potentially shows attractive neutronics performances on discharge burnup about 141GWd/ton with breeding ratio of 1.27, although rapid control rod scram and ACCS actuations are indispensable for core cooling under depressurization accident conditions.

JAEA Reports

Study on plant concept for Gas Cooled Fast Reactor

Moribe, Takeshi; Ikeda, Hirotsugu; Konomura, Mamoru

JNC-TY9400 2005-006, 30 Pages, 2005/06

JNC-TY9400-2005-006.pdf:1.4MB

In the

Oral presentation

Design study of double wall straight tube steam generator, 1; Design concept and structure

Kisohara, Naoyuki; Ikeda, Hirotsugu; Sato, Mitsuru*; Iitsuka, Toru*

no journal, , 

Safety, economy and public acceptance are required for commercialized FBR plant systems. Sodium heated steam generators of the FBRs must satisfy the plant safety and increase the plant availability and the safety impression to the society by decreasing the possibility of Na/water reaction as much as possible. For this purpose, the steam generators of the FBR provide double-wall-heat transfer tubes, and the basic structure of the SG was designed by taking account of Na/water reaction prevention and thermal hydraulic and structural viewpoints.

Oral presentation

Design study of double wall straight tube steam generator (SG), 3; Evaluation of thermo hydraulic analysis

Moribe, Takeshi; Sakai, Takaaki; Ikeda, Hirotsugu*; Yamada, Yumi*; Kurome, Kazuya*

no journal, , 

no abstracts in English

Oral presentation

Development of an fuel handling system for commercialized fast breeder reactor, 1; Conceptual design of fuel handling system

Usui, Shinichi; Chikazawa, Yoshitaka; Konomura, Mamoru; Tozawa, Katsuhiro*; Hori, Toru*; Toda, Mikio*; Ikeda, Hirotsugu

no journal, , 

no abstracts in English

Oral presentation

Development of an fuel handling system for commercialized fast breeder reactor, 2; Conceptual design of in-vessel fuel handling system

Chikazawa, Yoshitaka; Konomura, Mamoru; Usui, Shinichi; Tozawa, Katsuhiro*; Hori, Toru*; Toda, Mikio*; Ikeda, Hirotsugu*

no journal, , 

no abstracts in English

Oral presentation

Concept study on fuel handling system of commercialized sodium cooled reactor, 3; Study on ex-vessel sodium storage facility

Hori, Toru*; Tozawa, Katsuhiro*; Toda, Mikio*; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu*

no journal, , 

no abstracts in English

Oral presentation

Concept study on fuel handling system of commercialized sodium cooled reactor, 4; Study on spent fuel cleaning and water loading system

Tozawa, Katsuhiro*; Hori, Toru*; Toda, Mikio*; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu*

no journal, , 

no abstracts in English

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