Hydrogen isotope separation by cryogenic distillation

Yamanishi, Toshihiko

Purazuma, Kaku Yugo Gakkai-Shi, 92(1), p.21 - 25, 2016/01

AA2015-0316.pdf:0.93MB

http://www.jspf.or.jp/Journal/PDF_JSPF/jspf2016_01/jspf2016_01-21.pdf

In a fusion reactor, the hydrogen isotope separation system is required in the fuel cycle system to supply deuterium (D) and tritium (T) as its fuel. In ITER, 90% of T must be recycled through the isotope separation system. On the other hand; since the hydrogen (H) gas is finally exhausted to the environment, the T concentration in the H gas from the isotope separation system should be as low as reasonable achievable. Hence, the isotope separation system of a fusion reactor must have a large separation factor. The flow rate of the isotope separation system of a fusion reactor reaches to 300 mol/h. Only the cryogenic distillation method can meet the above conditions (large flow rate and separation factor) and is most likely used as a hydrogen separation system in a fusion reactor. In this chapter, several simulation methods and a set of experimental data of the cryogenic distillation columns are described in detail.

Development of fusion nuclear technologies at Japan Atomic Energy Research Institute

Seki, Masahiro; Yamanishi, Toshihiko; Shu, Wataru; Nishi, Masataka; Hatano, Toshihisa; Akiba, Masato; Takeuchi, Hiroshi; Nakamura, Kazuyuki; Sugimoto, Masayoshi; Shiba, Kiyoyuki; et al.

Fusion Science and Technology, 42(1), p.50 - 61, 2002/07

https://doi.org/10.13182/FST02-A212

Latest status on development of long-term fusion nuclear technologies at JAERI is overviewed. A tritium processing system for the ITER and DEMO reactors was designed and basic technologies for each component of this system was demonstrated successfully by an operation of the integrated system for one month. An ultra-violet laser with a wave length of 193 nm was found quite effective for removing tritium from in-vessel components of D-T fusion reactors. Blanket technologies have been developed for the Test Blanket Module of the ITER and for advanced blankets for DEMO reactors. This blanket is composed of LiTiO breeder pebbles and neutron multiplier Be pebbles, contained in a box structures made of a reduced activation ferritic steel F82H. Mechanical properties of F82H under neutron irradiation up to 50 dpa were obtained in a temperature range from 200 to 500C. Design of the International Fusion Materials Irradiation Facility (IFMIF) has been developed so as to obtain engineering data for candidate materials for DEMO reactors, under neutron irradiation up to 100-200 dpa.

Development of ceramic-coated lithium particles for tritium production tests in high temperature engineering test reactor

Yamashita, Kiyonobu; Sawa, Kazuhiro; Ando, Hiroei; ;

Nihon Genshiryoku Gakkai-Shi, 40(1), p.65 - 69, 1998/00

https://doi.org/10.3327/jaesj.40.65

no abstracts in English

Developments of tritium technology for next-step fusion devices under JAERI-DOE(LANL) collaboration

Naruse, Yuji; Okuno, Kenji; Yoshida, Hiroshi; Konishi, Satoshi; Anderson, J. L.*; Bartlit, J. R.*

Journal of Nuclear Science and Technology, 27(12), p.1081 - 1095, 1990/12

https://doi.org/10.3327/jnst.27.1081

no abstracts in English

Joint operation of TSTA under the collaboration between JAERI and DOE-LANL; An Integrated loop operation with 100-g tritium in July 1988

Okuno, Kenji; Enoeda, Mikio; ; ; Yoshida, Hiroshi; Naruse, Yuji; Anderson, J. L.*; Bartlit, J. R.*; Sherman, R. H.*; R.V.Carlson*; et al.

JAERI-M 90-028, 73 Pages, 1990/02

JAERI-M-90-028.pdf:1.51MB

no abstracts in English