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Journal Articles

Core and safety design for France-Japan common concept on sodium-cooled fast reactor

Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.

EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12

The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.

Journal Articles

Plant system study of France-Japan common concept on Sodium-cooled Fast Reactor

Kato, Atsushi; Yamamoto, Tomohiko; Ando, Masato; Chikazawa, Yoshitaka; Murakami, Hisatomo*; Oyama, Kazuhiro*; Kaneko, Fumiaki*; Higurashi, Koichi*; Chanteclair, F.*; Chenaud, M.-S.*; et al.

EPJ Nuclear Sciences & Technologies (Internet), 8, p.11_1 - 11_10, 2022/06

This paper provides an overview of plant system studies to establish a common technical view for Sodium-cooled Fast Reactor concept between France and Japan based on ASTRID600 and the new concept with downsized output called ASTRID150. One of important issues on a reactor structure design is to enhance seismic resistance to be tolerable against strong earthquake such that postulated in Japan. A concept of High Frequency Design is shared, and the design options related to HFD have been examined and design recommendations are established. In addition, this paper include results of studies for a steam generator, a decay heat removal system, a fuel handling system and a containment vessel.

Journal Articles

Progress in conceptual design of a pool-type sodium-cooled fast reactor in Japan

Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04

The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.

Journal Articles

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

Journal Articles

Coolability evaluation of debris bed on core catcher in a sodium-cooled fast reactor

Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05

Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.

Journal Articles

In-vessel thermal-hydraulics analyses of the ASTRID-600MWe reactor with STAR-CCM+ code to supply boundary conditions for mechanical evaluation

Onoda, Yuichi; Chikazawa, Yoshitaka; Nakamura, Hironori*; Barbier, D.*; Dirat, J.-F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

no abstracts in English

Journal Articles

ASTRID nuclear island design; Advances in French-Japanese joint team development of decay heat removal systems

Hourcade, E.*; Curnier, F.*; Mihara, Takatsugu; Farges, B.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1740 - 1745, 2016/04

In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving highlights of ASTRID DHRS current strategy. Focus is made on operating temperature diversification for in-vessel heat exchanger as well as core catcher coolability by an original features such as heat exchanger located within reactor cold pool, whose design was taken over by Japan team since 2014.

Journal Articles

Japan-France collaboration on the astrid program and sodium fast reactor

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.

Journal Articles

ASTRID, the SFR GENIV technology demonstrator project; Where are we, where do we stand for?

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.

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