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Kubo, Shigenobu; Payot, F.*; Yamano, Hidemasa; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; Shibata, Akihiro*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Rouault, J.*; Le Coz, P.*; Garnier, J.-C.*; Hamy, J.-M.*; Hayafune, Hiroki; Iitsuka, Toru*; Mochida, Haruo*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.832 - 837, 2015/05
The French and international industrial partners already joined the project from 2010 to 2013 and many others are also effective in the Research and Development in support of ASTRID. A new partnership is now effective on both topics with Japan. This collaboration on the ASTRID Program and Sodium Fast Reactor is now fully integrated in the ASTRID program organization. In addition a specific Joint Team, CEA, AREVA, JAEA, MHI and MFBR, has been created to follow specifically Japanese contribution and develop evaluations of a common interest to orientate future work and contribute to ASTRID options confirmation and be of an interest for the future Japanese Fast Breeder reactor.
Rouault, J.*; Abonneau, E.*; Settimo, D.*; Hamy, J.-M.*; Hayafune, Hiroki; Gefflot, R.*; Benard, R.-P.*; Mandement, O.*; Chauveau, T.*; Lambert, G.*; et al.
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.824 - 831, 2015/05
The Preconceptual Design phase of the ASTRID Project ended late 2012, the main goal was to evaluate innovative options. It is now followed by the AVP2 phase planned until the end of 2015 whose objectives are both to focus the design in order to finalize a coherent reactor outline and to finalize by December 2015 the Safety Option Report. The CEA acts as the industrial architect of the project. In 2014, Japan which participates now in the design studies and also in Research and Development in support of the ASTRID Project and VELAN are the latest partners to join the Project. The next important milestone is at the end of 2015 with the release by the Project team of a convincing and coherent Conceptual Design file.
Kikuchi, Hirohiko*; Mochida, Haruo*; Ide, Akihiro*; Iitsuka, Toru*; Hayafune, Hiroki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
Hishida, Masahiko; Konomura, Mamoru; Uchita, Masato; Iitsuka, Toru*; Kamishima, Yoshio*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), P. 5112, 2005/05
An innovative concept of medium-scale sodium-cooled modular reactor, named M-JSFR, has been created as based on the large-scale advanced loop type fast reactor concept. M-JSFR employs other concepts such as standardization and learning effects by designing as a modular plant and reduction of secondary loop number for the purpose of dissolving the scale-demerit. On this M-JSFR, some improvements are performed to overcome the weak point (strong chemical activity of sodium) of a sodium-cooled reactor and to achieve in-service inspection (ISI) and repair as easily as in light water reactors. Based on the ISI guidelines for light water reactors, the ISI procedures are reviewed reflecting the characteristics of M-JSFR. A guideline for ISI with the same grade of that of the light water reactors is established and major components subjected to ISI are selected. Moreover, suitable ISI procedures for each selected major component are proposed, and a plant concept amenable to ISI is studied. As the result of these studies, the construction cost of ISI&R reinforcement M-JSFR increases about 3% mainly because of the diameter expansion of reactor vessel.
Iitsuka, Toru; Oki, Yoshihisa; Kawashima, Shigeyo*; Nishimura, Motohiko; Isozaki, Tadashi; Kamide, Hideki
PNC TN9410 98-022, 58 Pages, 1998/03
Assessment of the maximum temperature and the position of the hot spot is the most important issues on the reactor safety when the local subchannel porous blockage is occurred. From these background, authors are going to perform a sodium experiment with 37-pin bundle test rig simulating the porous blockage, to understand the phenomena and acquire data for thermal-hydraulic analysis code validation. Before the execution of sodium test, one basic experiment and some using subchannel analysis code ASFRE-III had been done. The basic experiment was a water test to examine the pressure loss characteristics of the porous blockage. The pressure loss correlation derived from the water test was applied to the subsequent subchannel analysis of the 37-pin bundle sodium test rig. The analysis such predicted that the difference between the maximum temperature and the inlet temperature would be in propotion to the power to flow rate ratio, within the condition of the power=100
400 W/cm and the flow rate =200480 
/min. And it was also shown that the maximum subchannel temperature would not over the operational limit temperature 650 
C, if the power to flow rate ratio were kept lower than 0.75(W/cm)//min). The map was made to predict the maximum temperature from the experimental conditions.
Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.
PNC TN9410 96-279, 51 Pages, 1996/08
Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.
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.
