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Chikazawa, Yoshitaka; Aoto, Kazumi; Hayafune, Hiroki; Ono, Yushi; Kotake, Shoji; Toda, Mikio*; Ito, Takaya*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.426 - 435, 2011/05
Key technologies for Japan Sodium-cooled Fast Reactor (JSFR) has been evaluated. The hot vessel, two-loop cooling system using high chromium steel, integrated intermediate heat exchanger/pump component, highly reliable steam generator, natural circulation decay heat removal system and improved in-service inspection and repair capability have been confirmed to be feasible as development items for the next stage.
Devictor, N.*; Chikazawa, Yoshitaka; Saez, M.*; Rodriguez, G.*; Hayafune, Hiroki
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.664 - 673, 2011/05
CEA and JAEA intend to develop prototype or demonstration (proto/demo) of sodium-cooled fast reactors within these two decades. Common final goals of their respective programs are SFR commercialization. The target of commercialized SFR for both parties is basically consistent with the Generation IV goals. Due to industrial backgrounds and feedback of past/existing reactor experiences, ASTRID and JSFR have selected pool and loop configurations respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR). The analysis results showed that both concepts are technologically feasible and meet design goals. From the view point of collaboration, the present analysis has identified a wide range of collaborative items; they are described in the paper.
Aizawa, Kosuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.605 - 613, 2011/05
A conceptual design study of an advanced large-sized (1,500 MWe class) sodium-cooled fast reactor, JSFR, is in progress in the FaCT project in Japan. JSFR has adopted a selector-valve mechanism for a failed-fuel detection and location (FFDL) system. The selector-valve FFDL system identifies a failed fuel subassembly by sampling sodium from each fuel subassembly outlet and detecting fission product gas or delayed neutron precursors of fission products. One of the technologies which JSFR has adopted is an upper internal structure (UIS) with a radial slit. Because sampling nozzles cannot be set in the UIS slit, several sampling nozzles are installed around the slit so as to sample sodium from the failed fuel subassemblies under the UIS slit. In this study, a signal and noise detected by the delayed neutron detector have been calculated. On the basis of these results, appropriate operation patterns of the selector-valve FFDL system for JSFR have been constructed.
Yamano, Hidemasa; Kubo, Shigenobu*; Shimakawa, Yoshio*; Fujita, Kaoru; Suzuki, Toru; Kurisaka, Kenichi
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.728 - 740, 2011/05
Okubo, Tsutomu; Oki, Shigeo; Ogura, Masashi*; Okubo, Yoshiyuki*; Kotake, Shoji*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.479 - 486, 2011/05
A conceptual design study and related R&D on a commercial-base large-scale Japan Sodium-cooled Fast Reactor (JSFR) have been carried out in the framework of the Fast Reactor Cycle Technology development (FaCT) project. As a next generation plant, JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability and safety. This paper describes the current results of the ongoing conceptual design study on the JSFR core. The most important point in the core design is to achieve a high core average burn-up around 150 GWd/t, assuming the ODS steel utilization as the cladding material. Another design target for the breeding ratio is intended to have some flexibility and is set at from around 1.0 to 1.2 under the design philosophy of the compatible fuel assembly among them. Also, the fuel composition is considered to have some variation range based on the wide variety of the spent fuel composition expected to be treated during the LWR to FBR transition period. The core design study performed in the FaCT project has clarified the feasibility of the JSFR core concept, which is based on the high internal conversion ratio type core using a large fuel rod diameter around 10 mm and satisfies a number of design targets and requirements including ones mentioned above.
Am cross section with Monju reactor physics dataHazama, Taira; Takano, Kazuya; Kitano, Akihiro
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.1527 - 1535, 2011/05
The Japanese prototype fast breeder reactor Monju restarted its reactor physics test in May, 2010 after a 14-year interruption. The accumulation of Am due to the Pu decay during the interruption reaches 1.5wt% in average. An impact of the reactor physics data obtained in the restart core is investigated by the cross section adjustment technique with JENDL-3.3 and JENDL-4.0. Criticality data obtained before and after the interruption are applied. It is confirmed that Monju reactor physics data, when the two data are used together, effectively adjust Am capture cross sections. Consistent results are obtained among JENDL-3.3 after adjustment and JENDL-4.0 before and after the adjustment.
Shiotani, Hiroki; Ono, Kiyoshi; Heta, Masanori*; Yasumatsu, Naoto*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), 8 Pages, 2011/05
Comprehensive dynamic analyses of the typical Fast Reactor (FR) deployment scenarios with JSFR and related fuel cycle facilities developed in "FaCT: Fast Reactor Cycle Technology Development" project were conducted. In this study, combinations of analysis codes, which consist of two energy economic models (computable general equilibrium (CGE) model and energy system model) and dynamic analysis model for nuclear energy supply chain, were used for the dynamic analyses. As a result of the analysis, FR cycle should be deployed from 2040 to 2050 to curb the cumulative uranium resources consumption within the conventional uranium resources reported in Uranium 2009 by the OECD/NEA and the IAEA.
Oki, Shigeo; Meiliza, Y.; Kawashima, Katsuyuki; Okubo, Tsutomu
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.561 - 570, 2011/05
Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.1635 - 1643, 2011/05
In the situation that a variety of fuel composition is fed to FBRs during LWR-to-FBR transition stage, a special consideration on fuel reactivity and its loss by burnup is needed for each feeding fuel to keep criticality during operation period as prescribed. Suitable methods of determination of fuel enrichment will accomplish this without changing a pattern of fuel-loading (including number of loading fuel assembly). The choice of the method of enrichment determination affects the core characteristics which have to be controlled in core design. This paper describes some characteristics of the methods to determine an enrichment of fast reactor fuel from the core design points of view. Merits and demerits of these methods had been clarified in this study.
Ito, Kei; Koizumi, Yasuo*; Ohshima, Hiroyuki; Kawamura, Takumi*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.2085 - 2092, 2011/05
In the design study on a large-scale sodium-cooled fast reactor, the reactor vessel is compactified. However, such a reactor vessel induces higher coolant flows in the vessel and causes several thermal-hydraulics issues, e.g. gas entrainment (GE) phenomenon. To clarify the negative influences of the GE on the JSFR, not only the onset condition of the GE but also the entrained gas (bubble) flow rate has to be evaluated. In this study, the authors performs numerical simulations to investigate the entrained gas amount in a hollow vortex experiment. To simulate interfacial deformations accurately, a high-precision numerical simulation algorithm for gas-liquid two-phase flows is employed. As a result, the numerical simulation gives somewhat larger entrained gas flow rate than the experiment. However, both the numerical simulation and experiment show the entrained gas flow rates which are proportional to the outlet water velocity.
Okano, Yasushi; Sakai, Takaaki; Nakai, Ryodai
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.719 - 727, 2011/05
