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

https://doi.org/10.1051/epjn/2022040

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.

The evaluation of material base standard of ODS ferritic stainless steel core component for fast breeder reactors

Mizuta, Shunji; ;

JNC TN9400 2000-048, 28 Pages, 2000/04

JNC-TN9400-2000-048.pdf:0.64MB

ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(YO). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity

Development on extended burnup fuel technologies for practical high temperature gas-cooled reactors; Collaborative research with Kazakhstan

Ueta, Shohei; Shibata, Taiju; Aihara, Jun; Shaimerdenov, A.*; Dyussambayev, D.*; Takahashi, Masashi*; Kinoshita, Hideaki*; Gizatulin, S.*; Sakaba, Nariaki

no journal, , 

To develop the highly-qualified, mass-produced coated fuel particle in Japan which is supposed to be introduced to small modular commercial high temperature gas-cooled reactors (HTGRs), a dimensional specification of the fuel has been determined to attain three times higher burnup than that of the HTTR (High Temperature Engineering Test Reactor). Fabrication technologies of the fuel have been established in collaboration with Japanese nuclear fuel fabricator. As results on irradiation test and post irradiation examination at the Institute of Nuclear Physics in Kazakhstan via acceptances of two Regular Projects of ISTC (International Science and Technology Center), an excellent performance of the fuel under irradiation has been confirmed. Finally, technologies to extend the burnup for the highly-qualified, mass-produced HTGR fuel have been established first in the world.

Current status and issues on ODS tempered martensitic steel development for performance enhancement of advanced nuclear power system

Otsuka, Satoshi; Tanno, Takashi; Yano, Yasuhide; Fujita, Koji; Shizukawa, Yuta; Hashidate, Ryuta; Onizawa, Takashi; Kaito, Takeji; Ito, Chikara

no journal, , 

Implementation of fusion energy system and fast reactor cycle system requires the development of advanced materials resistant to the severe core environment where high-temperature and high-dose neutron irradiation are superposed. A lot of efforts have been made worldwide for research and development of oxide dispersion strengthened (ODS) steels with a variety of specification; Japan Atomic Energy Agency (JAEA) has focused on the development of 9Cr,11Cr-ODS tempered martensitic steel (TMS) for high-burnup fuel cladding tube of sodium-cooled fast reactor (SFR). This paper overviews the current status on 9Cr,11Cr-ODS TMS cladding tube development in JAEA, and discusses the cross-cutting issues in material development for advanced nuclear power systems.