Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Seismic qualification of piping systems by detailed inelastic response analysis, 4; Second round benchmark analyses with stainless steel piping component test

Watakabe, Tomoyoshi; Nakamura, Izumi*; Otani, Akihito*; Morishita, Masaki; Shibutani, Tadahiro*; Shiratori, Masaki*

Proceedings of the ASME 2017 Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07

https://doi.org/10.1115/PVP2017-65324

To provide a more rational seismic design, a new Code Case for seismic design of piping is now under development in the framework of JSME Nuclear Codes and Standards. The Code Case incorporates a dynamic elastic-plastic analysis procedure by employing finite element analysis as an alternative to the current design analysis method of elastic assumption. To confirm the applicability of inelastic response analysis, benchmark analyses have been conducted.

Chapter 11, Generation IV concepts: Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Morishita, Masaki

Handbook of Generation IV Nuclear Reactors, First Edition, p.283 - 307, 2016/00

https://doi.org/10.1016/C2014-0-01699-1

Handbook of Generation IV Nuclear Reactors is comprehensive resource on the research and advances in generation IV nuclear reactor concepts and the first edition is issued in 2016. The authors wrote the chapter 11: Generation IV nuclear reactor concepts: Japan, where developments activities on sodium cooled fast reactor in Japan are shown. Especially, concept of Japan sodium-cooled fast reactor (JSFR) is explained on the points of innovative technologies and safety enhancements after the TEPCO Fukushima Daiichi Nuclear Power Plant accident.

Development of structural codes for JSFR based on the system based code concept

Asayama, Tai; Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Nagae, Yuji; Takaya, Shigeru; Onizawa, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 6 Pages, 2014/07

This paper overviews the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR). Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.

Structural materials and code development for Japanese sodium-cooled fast reactors

Asayama, Tai; Nagae, Yuji; Wakai, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki

Proceedings of ASME Symposium on Elevated Temperature Application of Materials for Fossil, Nuclear, and Petrochemical Industries (ETAM2014), p.296 - 302, 2014/03

https://doi.org/10.1115/ETAM2014-1037

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were newly incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs.

Application of the system based code concept to the ASME code for liquid metal reactors

Asayama, Tai; Takaya, Shigeru; Morishita, Masaki

Proceedings of the ASME 2012 Pressure Vessels & Piping Conference (PVP 2012) (DVD-ROM), 6 Pages, 2012/07

https://doi.org/10.1115/PVP2012-78700

The System Based Code (SBC) which was initially proposed by Asada forms the basis of the development of a set of codes and standards in Japan that fully take advantage of the potential of liquid metal reactors (LMRs). The concept is also being elaborated within the American Society of Mechanical Engineers (ASME) as a promising way to improve current codes and standards which would allow significantly wider technical options with various advantages both in Sections III and XI by bridging them. This paper illustrates an envisioned scheme for code rules for new reactors based on the SBC concept that could be also implemented in ASME Boiler and Pressure Vessel Codes. The outline of technologies needed to materialize the concept such as "Break Size Limitation Analysis (BSLA)" and structural reliability evaluation are also mentioned. Recent discussions in ASME indicate that efforts on drafting code rules based on the SBC concept will soon take place in the ASME.

Main R&D issues for fast reactor structural design standard

Kasahara, Naoto; Nakamura, Kyotada; Morishita, Masaki; Shibamoto, Hiroshi; Inoue, Kazuhiko*

Nuclear Engineering and Design, 238(2), p.287 - 298, 2008/02

https://doi.org/10.1016/j.nucengdes.2006.09.018

For realization of economical and reliable Fast Reactor (FR) plants, the Japan Nuclear Cycle Development Institute (JNC) and the Japan Atomic Power Company (JAPC) are cooperating on the "Feasibility Study on Commercialized FR Cycle Systems". To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the "Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)" is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the "Refinement of Failure Criteria" need to be addressed for particular failure modes of FRs. Secondly, the development of "Guidelines for Inelastic Design Analysis" is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing "Guidelines for Thermal Load Modeling" for the design of FR components where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry.

Applicability of the system based code concept to the life-cycle design optimization of nuclear components

Asayama, Tai; Shibamoto, Hiroshi*; Otani, Tomomi*; Morishita, Masaki

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04

https://doi.org/10.1299/jsmeicone.2007.15._ICONE1510_430

The System Based Code concept intends to optimize margins for nuclear plant component throughout the service period, by offering methodologies to determine appropriate margins corresponding to various technical items, so that a total margin of the component is at a desired level. The authors have shown that this concept can successfully be applied to the design of relatively simple components of nozzles and end-caps. This paper explores the possibility of applying the concept of the System Based Code to more complex components whose design is directly connected to the design of the plant as a whole. As such example, a vessel equivalent to that of a regenerative heat exchanger of light water reactors was postulated. It was concluded that the concept could contribute to optimize the design of actual complicated components to the regions beyond the current codes and standards allow.

H ion beam acceleration in a single gap multi-aperture accelerator

Watanabe, Kazuhiro; Takayanagi, Tomohiro; Okumura, Yoshikazu; Hanada, Masaya; Inoue, Takashi; Kashiwagi, Mieko; Morishita, Takatoshi; Taniguchi, Masaki

JAEA-Technology 2005-002, 19 Pages, 2006/01

JAEA-Technology-2005-002.pdf:2.34MB

no abstracts in English

Experimental study on spatial uniformity of H ion beam in a large negative ion source

Hanada, Masaya; Seki, Takayoshi*; Takado, Naoyuki*; Inoue, Takashi; Morishita, Takatoshi; Mizuno, Takatoshi*; Hatayama, Akiyoshi*; Imai, Tsuyoshi*; Kashiwagi, Mieko; Sakamoto, Keishi; et al.

Fusion Engineering and Design, 74(1-4), p.311 - 317, 2005/11

https://doi.org/10.1016/j.fusengdes.2005.06.259

no abstracts in English