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Fabrication routes for advanced first wall design alternatives

Rieth, M. ORCID iD icon 1; Dürrschnabel, M. ORCID iD icon 1; Bonk, S. 1; Antusch, S. 1; Pintsuk, G.; Aiello, G.; Henry, J.; Carlan, Y. de; Ghidersa, B.-E. 2; Neuberger, H. 2; Rey, J. 2; Zeile, C. 2; De Wispelaere, N.; Simondon, E. 1; Hoffmann, J. 1
1 Karlsruher Institut für Technologie (KIT)
2 Institut für Neutronenphysik und Reaktortechnik (INR), Karlsruher Institut für Technologie (KIT)

Abstract (englisch):

In future nuclear fusion reactors, plasma facing components have to sustain specific neutron
damage. While the majority of irradiation data provides a relatively clear picture of the
displacement damage, the effect of helium transmutation is not yet explored in detail.
Nevertheless, available results from simulation experiments indicate that 9%-chromium steels
will reach their operating limit as soon as the growing helium bubbles extent a critical size. At
that point, the material would most probably fail due to grain boundary embrittlement. In this
contribution, we present a strategy for the mitigation of the before-mentioned problem using
the following facts. (1) The neutron dose and related transmutation rate decreases quickly
inside the first wall of the breeding blankets, that is, only a plasma-near area is extremely
loaded. (2) Nanostructured oxide dispersion strengthened (ODS) steels may have an enormous
trapping effect on helium, which would suppress the formation of large helium bubbles for a
much longer period. (3) Compared to conventional steels, ODS steels also provide improved
irradiation tensile ductility and creep strength. ... mehr

Verlagsausgabe §
DOI: 10.5445/IR/1000139279
Veröffentlicht am 25.10.2021
DOI: 10.1088/1741-4326/ac2523
Zitationen: 5
Web of Science
Zitationen: 5
Zitationen: 6
Cover der Publikation
Zugehörige Institution(en) am KIT Institut für Angewandte Materialien – Werkstoffkunde (IAM-WK)
Publikationstyp Zeitschriftenaufsatz
Publikationsmonat/-jahr 11.2021
Sprache Englisch
Identifikator ISSN: 0029-5515, 1741-4326
KITopen-ID: 1000139279
HGF-Programm 31.13.05 (POF IV, LK 01) Neutron-Resistant Structural Materials
Erschienen in Nuclear fusion
Verlag International Atomic Energy Agency (IAEA)
Band 61
Heft 11
Seiten 116067
Vorab online veröffentlicht am 14.10.2021
Schlagwörter blanket first wall, oxide dispersion strengthened (ODS) steel, high heat flux test, helium cooling loop, materials technology, dissimilar joints, diffusion bonding
Nachgewiesen in Web of Science
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