Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, K.; Kasada, Ryuta; Wakai, Eiichi; Serizawa, Hisashi; Kawahito, Yousuke; Jitsukawa, Shiro; Kimura, Akihiko; Kohno, Yutaka; Kohyama, Akira; Katayama, Seiji; Mori, Haruhiko; Nishimoto, Kazutoshi; Klueh, R.L.; Sokolov, Mikhail A.; Stoller, R.E.; Zinkle, S.J.

doi:10.1016/j.fusengdes.2008.07.024

Cited by 84 publications

(25 citation statements)

References 21 publications

(28 reference statements)

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Reduced activation ferritic-martensitic steel F82H is one of candidate materials for the first wall in fusion reactor [1,2], and many studies on this steel has been reported so far [3]. The first wall in the reactor receives various particles, such as neutron and helium ash, and then the damages are introduced into the surface.…”

Section: Introductionmentioning

confidence: 99%

Deuterium retention and desorption behavior of reduced activated ferritic steel with surface damage due to high energy helium ion irradiation

Yamauchi

Gotoh

Nobuta

et al. 2010

Fusion Engineering and Design

View full text Add to dashboard Cite

Reduced activation ferritic-martensitic steel F82H was irradiated with 5 keV helium ions to introduce the surface damage. After the irradiation, the irradiated F82H was continuously exposed to 1.7 keV deuterium ions. The effects of the damages introduced by the helium ion irradiation on deuterium retention and desorption behavior of F82H were investigated. The HD desorption had a peak at around 500 K, which was shifted to the high temperature region when the helium ion fluence increased. The decrease in HDO desorption at the high fluence region was observed. This might be owing to the oxygen depletion on the surface by the sputtering. The amount of retained deuterium of the F82H gradually increased with the helium ion fluence and saturated at 7.2 dpa.

show abstract

Section: Introductionmentioning

confidence: 99%

Deuterium retention and desorption behavior of reduced activated ferritic steel with surface damage due to high energy helium ion irradiation

Yamauchi

Gotoh

Nobuta

et al. 2010

Fusion Engineering and Design

View full text Add to dashboard Cite

show abstract

“…These thermo-mechanical and hydraulic tests have proved the integrity of the TBM structure and fabrication technologies [47]. Several first-wall irradiation tests for base materials and bonded structures demonstrated that the performance was well within an acceptable level [48].…”

Section: Tbm Developmentmentioning

confidence: 99%

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

Matsuda

Tobita²

2013

Journal of Nuclear Science and Technology

View full text Add to dashboard Cite

An international joint project of fusion experimental reactor, the ITER (International Thermonuclear Experimental Reactor), is reviewed in view of long-range fusion energy research and development (R&D). Its purpose, goal, evolution, and the present construction status are briefly reviewed. While the ITER is a core machine in the present stage, generation of electricity is a role of the next-step fusion demonstration power plant "DEMO." The status of designs and technology R&D for DEMO are also reviewed.

show abstract

“…Joining and inspection technologies are also important for the fabrication of the DEMO blanket, because the blanket design is a complex box structure with numerous joints. The weld joints of high-Cr ferritic steels, especially in the heat-affected zone (HAZ), are known as the weakest parts of all the structures [4]. Thus, it is important to optimize the weld technologies in order to minimize the HAZ region and understand the mechanisms of various effects that appear at the weld joints, so that one can estimate the impact of these effects under irradiation.…”

Section: Japanese Randd On Rafm Steel In Ba Activitymentioning

confidence: 99%

Technical issues related to the development of reduced-activation ferritic/martensitic steels as structural materials for a fusion blanket system

Tanigawa

Shiba

Sakasegawa

et al. 2011

Fusion Engineering and Design

View full text Add to dashboard Cite

Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

Cited by 84 publications

References 21 publications

Deuterium retention and desorption behavior of reduced activated ferritic steel with surface damage due to high energy helium ion irradiation

Deuterium retention and desorption behavior of reduced activated ferritic steel with surface damage due to high energy helium ion irradiation

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

Technical issues related to the development of reduced-activation ferritic/martensitic steels as structural materials for a fusion blanket system

Contact Info

Product

Resources

About