[HTML][HTML] Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review
A Bhattacharya, SJ Zinkle, J Henry… - Journal of Physics …, 2022 - iopscience.iop.org
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 …
steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The …
[HTML][HTML] Current state and prospect on the development of advanced nuclear fuel system materials: A review
The intricate balance between reactor economics and safety necessitates the emergence of
new and advanced nuclear systems and, very importantly, advanced materials, which can …
new and advanced nuclear systems and, very importantly, advanced materials, which can …
A review of the irradiation evolution of dispersed oxide nanoparticles in the bcc Fe-Cr system: Current understanding and future directions
Thus far, a number of studies have investigated the irradiation evolution of oxide
nanoparticles in bcc Fe-Cr based oxide dispersion strengthened (ODS) alloys. But given the …
nanoparticles in bcc Fe-Cr based oxide dispersion strengthened (ODS) alloys. But given the …
Radiation-induced Ostwald ripening in oxide dispersion strengthened ferritic steels irradiated at high ion dose
Oxide dispersion strengthened (ODS) ferritic steels are considered promising candidates as
cladding tubes for Generation IV nuclear reactors. In such reactors, irradiation damage can …
cladding tubes for Generation IV nuclear reactors. In such reactors, irradiation damage can …
Oxide dispersion-strengthened/ferrite-martensite steels as core materials for Generation IV nuclear reactors
S Ukai, S Ohtsuka, T Kaito, Y De Carlan, J Ribis… - Structural materials for …, 2017 - Elsevier
Oxide dispersion-strengthened (ODS) steels are the most promising candidate materials for
fuel cladding of Generation IV nuclear reactors. The progress and current status for …
fuel cladding of Generation IV nuclear reactors. The progress and current status for …
Microstructural changes and void swelling of a 12Cr ODS ferritic-martensitic alloy after high-dpa self-ion irradiation
A dual-phase 12Cr oxide-dispersion-strengthened (ODS) alloy, with improved corrosion and
oxidation resistance exhibits promising void swelling resistance and microstructural stability …
oxidation resistance exhibits promising void swelling resistance and microstructural stability …
Temperature dependent dispersoid stability in ion-irradiated ferritic-martensitic dual-phase oxide-dispersion-strengthened alloy: Coherent interfaces vs. incoherent …
In this study, the microstructure of a 12Cr ferritic-martensitic oxide-dispersion-strengthened
(ODS) alloy is studied before and after Fe ion irradiation up to 200 peak displacements per …
(ODS) alloy is studied before and after Fe ion irradiation up to 200 peak displacements per …
Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys
In this research, innovative thermal spray deposition (Process I) and conventional hot
extrusion processing (Process II) methods have been used to produce thin walled tubing (∼ …
extrusion processing (Process II) methods have been used to produce thin walled tubing (∼ …
Depositing laser-generated nanoparticles on powders for additive manufacturing of oxide dispersed strengthened alloy parts via laser metal deposition
R Streubel, MB Wilms, C Doñate-Buendía… - Japanese Journal of …, 2018 - iopscience.iop.org
We present a novel route for the adsorption of pulsed laser-dispersed nanoparticles onto
metal powders in aqueous solution without using any binders or surfactants. By electrostatic …
metal powders in aqueous solution without using any binders or surfactants. By electrostatic …
The comparison of microstructure and nanocluster evolution in proton and neutron irradiated Fe–9% Cr ODS steel to 3 dpa at 500 C
MJ Swenson, JP Wharry - Journal of nuclear materials, 2015 - Elsevier
A model Fe–9% Cr oxide dispersion strengthened (ODS) steel was irradiated with protons or
neutrons to a dose of 3 displacements per atom (dpa) at a temperature of 500° C, enabling a …
neutrons to a dose of 3 displacements per atom (dpa) at a temperature of 500° C, enabling a …