Low-temperature radiation embrittlement of copper alloys

The effect of low-temperature (T{sub irr} less than 0.3. T{sub melt}) irradiation on the tensile properties of copper and precipitation-hardened (PH) and dispersion-strengthened (DS) copper alloys was investigated. Samples were irradiated with fission neutrons at temperatures of 80 to 200{degrees}C to doses of 0.6 to 5 dpa. Irradiation at temperatures <150{degrees}C resulted in significant hardening and accompanying embrittlement in all of the materials. By comparing the present results with literature data, it is concluded that severe radiation embrittlement occurs in copper alloys irradiated at temperatures {le}I00{degrees}C for doses above {approximately} 0.01 to 0.1 dpa. On the other hand, irradiation at temperatures above 150{degrees}C causes only moderate embrittlement for doses up to {approximately}5 dpa. It is recommended that the minimum operating temperature for copper alloys intended for structural applications in fusion energy systems should be 150{degrees}C, unless uniform elongations <I% can be accomodated in the design

Exploring electronic effects on the partitioning of actinides(III) from lanthanides(III) using functionalised bis-triazinyl phenanthroline ligands

The first examples of 4,7-disubstituted 2,9-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-1,2,4-benzo-triazin-3-yl)-1,10- phenanthroline (CyMe4 -BTPhen) ligands are reported herein. Evaluating the kinetics, selectivity and stoichiometry of actinide(III) and lanthanide(III) radiotracer extractions has provided a mechanistic insight into the extraction process. For the first time, it has been demonstrated that metal ion extraction kinetics can be modulated by backbone functionalisation and a promising new CHON compliant candidate ligand with enhanced metal ion extraction kinetics has been identified. The effects of 4,7- functionalisation on the equilibrium metal ion distribution ratios are far more pronounced than those of 5,6-functionalisation. The complexation of Cm(III) with two of the functionalised ligands was investigated by TRLFS and, at equilibrium, species of 1:2 [M:L] stoichiometry were observed exclusively. A direct correlation between the ELUMO-EHOMO energy gap and metal ion extraction potential is reported, with DFT studies reaffirming experimental findings

Perspectives on multiscale modelling and experiments to accelerate materials development for fusion

Prediction of material performance in fusion reactor environments relies on computational modelling, and will continue to do so until the first generation of fusion power plants come on line and allow long-term behaviour to be observed. In the meantime, the modelling is supported by experiments that attempt to replicate some aspects of the eventual operational conditions. In 2019, a group of leading experts met under the umbrella of the IEA to discuss the current position and ongoing challenges in modelling of fusion materials and how advanced experimental characterisation is aiding model improvement. This review draws from the discussions held during that workshop. Topics covering modelling of irradiation-induced defect production and fundamental properties, gas behaviour, clustering and segregation, defect evolution and interactions are discussed, as well as new and novel multiscale simulation approaches, and the latest efforts to link modelling to experiments through advanced observation and characterisation techniques.MRG, SLD, and DRM acknowledge funding by the RCUK Energy Programme [grant number EP/T012250/1]. Part of this work has been carried out within the framework of the EUROFusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant Agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. JRT acknowledges funding from the US Department of Energy (DOE) through grant DE-SC0017899. ZB, LY,BDW, and SJZ acknowledge funding through the US DOE Fusion Energy Sciences grant DE-SC0006661ZB, LY and BDW also were partially supported from the US DOE Office of Science, Office of Fusion Energy Sciences and Office of Advanced Scientific Computing Research through the Scientific Discovery through Advanced Computing (SciDAC) project on Plasma-Surface Interactions. JMa acknowledges support from the US-DOEs Office of Fusion Energy Sciences (US-DOE), project DE-SC0019157. Pacific Northwest National Laboratory is operated by Battelle Memorial Institute for the US Department of Energy (DOE) under contract DE-AC05-76RL01830. YO and YZ were supported as part of the Energy Dissipation to Defect Evolution (EDDE), an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Basic Energy Sciences under contract number DE-AC05-00OR22725. TS and TT are supported by JSPS KAKENHI Grant Number 19K05338

Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments

Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M(23)C(6) precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments

Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

a b s t r a c t The Japan-USA collaborative program, JUPITER-II, has made significant progress in a research program titled &quot;The irradiation performance and system integration of advanced blanket&quot; through a six-year plan for [2001][2002][2003][2004][2005][2006]. The scientific concept of this program is to study the elemental technology in macroscopic system integration for advanced fusion blankets based on an understanding of the relevant mechanics at the microscopic level. The program has four main research emphases: (1) Flibe molten salt system: Flibe handling, reduction-oxidation control by Be and Flibe tritium chemistry; thermofluid flow simulation experiment and numerical analysis. (2) Vanadium /Li system: MHD ceramics coating of vanadium alloys and compatibility with Li; neutron irradiation experiment in Li capsule and radiation creep. (3) SiC/He system: Fabrication of advanced composites and property evaluation; thermomechanics of SiC system with solid breeding materials; neutron irradiation experiment in He capsule at high temperatures. (4) Blanket system modeling: Design-based integration modeling of Flibe system and V/Li system; multiscale materials system modeling including He effects. This paper describes the perspective of the program including the historical background, the organization and facilities, and the task objectives. Important recent results are reviewed

Thermal Evolution of the Proton Irradiated Structure in Tungsten–5 wt% Tantalum

We have monitored the thermal evolution of the proton irradiated structure of W–5 wt% Ta alloy by in-situ annealing in a transmission electron microscope at fusion reactor temperatures of 500–1300 °C. The interstitial-type a/2 dislocation loops emit self-interstitial atoms and glide to the free sample surface during the early stages of annealing. The resultant vacancy excess in the matrix originates vacancy-type a/2 dislocation loops that grow by loop and vacancy absorption in the temperature range of 600–900 °C. Voids form at 1000 °C, by either vacancy absorption or loop collapse, and grow progressively up to 1300 °C. Tantalum delays void formation by a vacancy-solute trapping mechanism