High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

https://doi.org/10.1016/C2018-0-05383-9

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Development of PRW welding technology for 9Cr-ODS cladding tube

Seki, Masayuki; Kihara, Yoshiyuki; Kaito, Takeji; Tsukada, Tatsuya*; Motoki, Kazuhiko*; Hirako, Kazuhito*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Oxide Dispersion Strengthened (ODS) steel has been developed as an advanced fuel cladding tube for sodium cooled fast reactors in Japan Atomic Energy Agency (JAEA) to attain the target burn up of 150 GWd/t in the bundle average because of its excellent swelling resistance and high mechanical strength in high temperature. If conventional TIG welding is applied to the ODS welding, it is difficult to obtain necessary mechanical strength at the weld zone because of the formation of porosity. It is formed by the argon bubbles which initially dissolve in the matrix and grow up under the high temperature during welding. Therefore JAEA has been conducted the development of pressurized resistance welding (PRW) technology for ODS cladding tube, which is one of the solid state welding methods. This paper describes in the development of PRW technology, an ultrasonic test method for detecting weld defects, the result of machine strength measurement examination in weld part and the result of fuel pin irradiation examination using nuclear reactor.

Conceptual design for a large-scale Japan sodium-cooled fast reactor, 3; Core design in JSFR

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

http://www.icapp.ans.org/icapp11/program/abstracts/11345.pdf

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.

Carbon-14 analysis in solidified product of non-metallic solid waste by a combination of alkaline fusion and gaseous CO trapping

Ishimori, Kenichiro; Kameo, Yutaka; Matsue, Hideaki; Oki, Yoshiyuki*; Nakashima, Mikio; Takahashi, Kuniaki

Applied Radiation and Isotopes, 69(2), p.506 - 510, 2011/02

https://doi.org/10.1016/j.apradiso.2010.10.014

In order to establish a simple and rapid analytical method for C in solidified products made from non-metallic low-level radioactive solid wastes by melting treatment, a radiochemical analysis in combination with alkaline fusion as a sample decomposition method was examined. A simulated solidified product containing C, which was prepared by using nuclear reaction N(n, p)C with thermal neutron irradiation, was analyzed by the present method to compare with a conventional radiochemical analysis using oxidizing combustion. The reproducible and quantitative recovery of C from the simulated solidified product indicates that the present method is superior and more efficient for C analysis in solidified products than the conventional method using oxidizing combustion.

Preparation of reference material for solidified products made from radioactive miscellaneous wastes by melting treatment; Reference material for solidified product containing -ray emitting nuclides (Joint research)

Ishimori, Kenichiro; Oki, Keiichi; Takaizumi, Hirohide; Kameo, Yutaka; Oki, Yoshiyuki*; Nakashima, Mikio

JAEA-Technology 2007-065, 20 Pages, 2008/01

JAEA-Technology-2007-065.pdf:1.4MB

In order to prepare a reference material which is used for radiochemical analysis of solidified products made from non-metallic miscellaneous low level radioactive solid wastes by melting in Nuclear Science Research Institute, Japan Atomic Energy Agency, the preparation method of the reference material was investigated. Under the optimum melting conditions obtained in this report, the reference material containing Np, Am and Cm as -ray emitting nuclides was successfully prepared. From radiochemical analysis of the reference material, the radioactive concentration of respective nuclides was determined to be 0.1880.001 Bq/g for Np, 0.3680.004 Bq/g for Am, 0.4020.010 Bq/g for Cm.

Analyses of heat load in ITER NBI duct and neutron streaming through pressure relief line

Sato, Satoshi; Yamauchi, Michinori; Nishitani, Takeo; Ioki, Kimihiro; Iida, Hiromasa; Kataoka, Yoshiyuki

JAEA-Technology 2006-032, 91 Pages, 2006/03

JAEA-Technology-2006-032.pdf:12.8MB

no abstracts in English

Coolant flow of HTTR; Numerical study for flow pattern of coolant under core support plate

Inagaki, Yoshiyuki; Fujimoto, Nozomu; Motoki, Yasuo; Iyoku, Tatsuo; Maruyama, So; Shiozawa, Shusaku

JAERI-M 90-223, 30 Pages, 1990/12

JAERI-M-90-223.pdf:0.78MB

no abstracts in English

Simplification of procedure for radiochemical analysis of solidified products, 6; Separation method of Se

Kameo, Yutaka; Katayama, Atsushi; Nakashima, Mikio; Oki, Yoshiyuki*; Isogai, Keisuke*

no journal, , 

no abstracts in English

Crystallization and a preliminary neutron diffraction experiment of an antifreeze protein from Notched-fin eelpout

Ohara, Takashi; Adachi, Motoyasu; Tamada, Taro; Kuroki, Ryota; Kondo, Hidemasa*; Nishimiya, Yoshiyuki*; Tsuda, Sakae*

no journal, , 

no abstracts in English

Conceptual design study on core toward the demonstration reactor of JSFR, 1; Project summary

Okubo, Tsutomu; Oki, Shigeo; Ogura, Masashi*; Okubo, Yoshiyuki*

no journal, , 

Aiming at realization of the demonstration reactor for the Japan Sodium-cooled Fast Reactor (JSFR) around 2025, the conceptual design study has been performed on 750 or 500 MWe core. This presentation gives the present status. In the conceptual design of the demonstration reactor, design targets, requirements and conditions are set as the same as those of the 1500 MWe commercial reactor. The design of the fuel assembly and fuel pin are also the same as in the commercial reactor, based on the consideration to demonstrate them. Fuel composition covers MA bearing fuel supposed in the commercial reactor and the highly decontaminated U-Pu fuel considering the fuel cycle circumstances around 2025. The study has shown that the core concept satisfying given design condition will be possible, as well as the range of core performance change based on the fuel composition range.