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

Abstract

Reduced activation ferritic-martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. Performance of these steels will deteriorate in-service due to neutron damage and transmutation-induced gases like helium/hydrogen at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the major irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (i) 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, (ii) inter-mediate temperature cavity swelling, (iii) the effect of helium on LTHE and cavity swelling, (iv) irradiation creep and (v) tritium management issues are reviewed. The potential causes of LTHE are discussed that highlights need for advanced characterization techniques. Mechanical properties including tensile/Charpy impact of RAFM and ODS steels are compared to show that current generation of ODS steels also suffer from LTHE, and they can show irradiation hardening up to high temperatures, ~400-500 °C. To minimize this, future ODS steel development for FW/B specific application should target materials with lower Cr concentration (to minimize α’) and minimize 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 synergistic effects of helium on thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex including stresses due to 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 multi-scale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility

Nuclear data for fusion: inventory validation successes and future needs

“…Such a predicted difference, if validated, could have dramatic implications for any carbon compound's potential use in a fusion reactor-either the material lifetime will be severely limited due to helium-induced swelling or embrittlement, or the material will require advanced microstructural engineering to allow it to accommodate high gas production without failure. In any case, the observations here confirm that there must be an effort to reinstate the (n,n ′ 2α) for 12 C into the next releases of libraries like TENDL, and also that there is a need for further experiments to evaluate this potentially highly impactful reaction channel-the data points are highly scattered in figure 6.…”

“…In figure 6, EXFOR cross section data associated with the typical (n,α) channel for He production is shown for the primary stable isotope of C, 12 C (98.93% of natural carbon). Also shown, is data attributed to an alternative, more exotic nuclear reaction channel, which involves neutron capture followed by the break-up of 12 C into three α-particles and a residual neutron. Typically designated as (n,n ′ 2α) (the third α particle of 4 He nucleus is the remaining residual in this convention), the figure shows that there is experimental data that provides strong evidence of a non-negligible cross section for this channel at and around the 14 MeV neutron energies of the deuterium-tritium fusion reaction.…”

“…Figure6. Helium (α-particle) production cross sections on12 C, comparing the curves from the TENDL-2021[23] with those from EAF-2003[56]. EXFOR experimental cross section data points for the two different reaction channels (n,α) and (n,n ′ 2α) are also shown.…”

APT characterization of irradiation effects on MX phase in reduced-activation ferritic/martensitic steels

“…The reason for the bigger size of MX precipitates in Eurofer 97 is still unclear. However, it could be one reason for that RAFM steels show worse thermal creep properties than many present-day 9% Cr steels [47] , such as the 9Cr3W3CoVNbBN steel [48] , 10%Cr martensitic steels [49] , and 9Cr-1Mo martensitic steel [50] .…”

Monte Carlo calculation of absorbed dose under MeV proton irradiation

“…For instance, during a typical neutron irradiation experiment in a thermal test reactor, the end-of-life damage is 3-5dpa/year, likewise a fast reactor gives 20dpa/year. The average of end-of-life damage for components of a BWR core is 10 dpa, for PWR it is 80 dpa and for Advanced Fast Reactor it is 200 dpa [22][23][24][25] The computational simulation used was made with the FORTRAN programming language in which is only compatible with a Linux operating system. The simulation takes into account target material properties and several incident beam parameters.…”