Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Bhattacharya, Arunodaya; Zinkle, Steven J.; Henry, Jean; Levine, Samara M.; Edmondson, Philip D.; Gilbert, Mark R.; Tanigawa, Hiroyasu; Kessel, Charles E.

doi:10.1088/2515-7655/ac6f7f

Title: Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Journal Article · Wed Jul 06 00:00:00 EDT 2022 · JPhys Energy

DOI:https://doi.org/10.1088/2515-7655/ac6f7f· OSTI ID:1876325

^[1]; Zinkle, Steven J. ^[2]; Henry, Jean ^[3]; Levine, Samara M. ^[4];

^[1];

^[5]; Tanigawa, Hiroyasu ^[6]; Kessel, Charles E. ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States)
Univ. Paris-Saclay, Gif-sur-Yvette (France)
Univ. of Tennessee, Knoxville, TN (United States)
United Kingdom Atomic Energy Authority (UKAEA), Abingdon (United Kingdom)
National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori (Japan)

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

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Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Science (SC), Fusion Energy Sciences (FES); USDOE Laboratory Directed Research and Development (LDRD) Program; RCUK Energy Programme

Grant/Contract Number:: AC05-00OR22725; EP/T012250/1

OSTI ID:: 1876325

Journal Information:: JPhys Energy, Vol. 4, Issue 3; ISSN 2515-7655

Publisher:: IOP PublishingCopyright Statement

Country of Publication:: United States

Language:: English

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