Transmutation Effects in Fusion Reactor Materials: Critical Challenges & Path Forward: Helium, Tritium and Hydrogen Effects III
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Arunodaya Bhattacharya, Oak Ridge National Laboratory; Steven Zinkle, University of Tennessee; Philip Edmondson, The University Of Manchester; Aurelie Gentils, Université Paris-Saclay; David Sprouster, Stony Brook University; Takashi Nozawa, National Institutes for Quantum and Radiological Science and Technology (QST); Martin Freer, University of Birmingham
Wednesday 2:00 PM
March 22, 2023
Room: 27B
Location: SDCC
Session Chair: Yan-Ru Lin, Oak Ridge National Lab; Mark Gilbert, UK Atomic Energy Authority
2:00 PM Invited
Transmutation Effects in Fine Grained Tungsten: Gas Behavior and the Role of Grain Boundaries: Jason Trelewicz1; 1Stony Brook University
In tungsten, which has emerged as the primary armor material for the divertor in ITER and a likely candidate for early demonstration reactors, neutron irradiation promotes embrittlement and property degradation with enhanced production of gaseous (H/He) and solid (Re/Os) transmutation products under the high-energy fusion neutron spectrum. In this presentation, these effects are reviewed with an emphasis on research opportunities in grain boundary engineered tungsten. A focused set of experiments on the implications of helium gas production for the mechanical behavior is described, employing fine grained tungsten specifically to probe the role of grain boundaries. We show that the formation of grain boundary cavities is correlated to a reduction in hardness, which is recovered with increasing fluence due to classical irradiation hardening. Insights from molecular dynamic simulations demonstrate that the observed softening can be attributed to stress-assisted grain boundary bubble coalescence and enhanced strain accommodation within the grain boundaries.
2:40 PM
Interaction of Hydrogen/Helium with Grain Boundaries and Dislocations in Tungsten: Nithin Mathew1; Enrique Martinez2; Blas Uberuaga1; Danny Perez1; 1Los Alamos National Laboratory; 2Clemson University
Tungsten (W) is the primary candidate for divertor material in future fusion reactors. In addition to intake from the fuel, nuclear transmutation reactions lead to formation of H and He in the divertor. Interaction of these H/He with extended defects such as dislocations and grain boundaries (GBs) can have a significant effect on the microstructure and mechanical properties of W divertor. We will present predictions from direct-/accelerated- molecular dynamics simulations on the interactions between H/He with symmetric tilt GBs and dislocations. Segregation of H is found to result in a complex, structure dependent effect on GB mobility. Assuming a disconnection-based mechanism for GB motion, it is shown that the conventional picture of solute-drag dominated Arrhenius kinetics is incomplete. He bubble growth in the vicinity of dislocations results in the formation sessile and glissile loops which, in turn, leads to complex reactions with the existing dislocation and modification of its character.
3:00 PM
Helium Production in Irradiated Low-temperature Solder Candidates for Novel Fusion Magnet Cables: Christopher Reis1; 1University of California, Berkeley
As momentum behind commercially viable nuclear fusion builds, solder-impregnated REBCO cables, such as MIT’s VIPER cables, have gained favor for their electrical, mechanical, and thermal properties. However, solder in these novel cables would be subject to the fusion-reaction neutrons and thus helium bubble production. Since in-service conditions require cryogenic temperatures for the superconducting magnets, the irradiation and subsequent bubble production would occur around 20-30K. As these magnets are then brought to room temperature for maintenance the bubbles would undergo significant expansion, potentially leading to cable fracturing from high pressure bursting and would result in an electromechanically compromised system. This study proposes to investigate the effects of helium production as a function of fluence by first simulating the He production expected using MCNP/FISPACT-II, then using cryo-helium implantation at 77K to replicate the effect of the irradiation on candidate solders before subsequently raising to room temperature for mechanical testing.
3:20 PM
Accurate Fe-He Machine Learning Potential for Studying Helium Effects in Ferritic Steels: Krishna Pitike1; Wahyu Setyawan1; 1Pacific Northwest National Laboratory
Nanostructured Ferritic Alloys (NFAs) are being actively investigated for their potential application as structural materials in advanced fusion reactors. Due to the lack of fusion test reactors, a predictive mesoscale model is required to understand radiation damage in NFAs, including helium bubble accumulation effects. Machine learning interatomic potentials (MLPs) have shown promising results due to their high accuracy for a fractional computational cost, and high adaptability towards complex chemical environments, such as Fe-He-H-(YTO) in NFAs. Here, we develop a Fe-He potential, based on ≈10,000 atomic configurations sampled using DFT. The developed MLP predicts the bulk properties of BCC-Fe, and thermodynamics of helium bubbles accurately in the iron matrix. E.g., the mean average error (MAE) of He binding energies with HenV (Hen) bubbles (clusters) estimated using MLP is ≈4 times smaller than that of classical potentials. The current Fe-He MLP can be further developed to include all chemical interactions in NFAs.
3:40 PM Break
4:00 PM
Behavior of Helium Cavities in Ion-irradiated Ductile-Phase-Toughened Tungsten: Weilin Jiang1; Libor Kovarik1; Karen Kruska1; Dalong Zhang1; Dongsheng Li1; Tamas Varga1; Wahyu Setyawan1; 1Pacific Northwest National Laboratory
Ductile-phase-toughened tungsten (DPT W) composites, such as W-Ni-Fe DPT W, have been investigated as a candidate material for the plasma-facing components of fusion reactors. The DPT W in this study consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. Sequential irradiation with Ni⁺ and He⁺ ions was performed at 973 K to emulate high-energy neutron irradiation and transmutation effects in the material. Larger He cavities with a lower number density are observed in NiFeW than W. Cavities are aggregated preferentially along the NiFeW/W interphase boundary. Convergent-beam scanning transmission electron microscopy was applied to image He cavities at an atomic-scale resolution. The depth profiles of the cavity size/volume and number density are determined. A quantitative analysis of the He atomic density and pressure in the cavities with sizes down to only ~3 nm has been achieved using electron energy loss spectroscopy.
4:20 PM
Machine-learned Interatomic Potential Development for H Trapping in ZrC Strengthened W: Ember Sikorski1; Mary Alice Cusentino1; Megan McCarthy1; Julien Tranchida2; Mitchell Wood1; Aidan Thompson1; 1Sandia National Laboratories; 2CEA Cadarache
While tungsten is a leading candidate material for plasma-facing components, it has a high ductile-to-brittle transition temperature. Additionally, W may undergo recrystallization and grain growth at fusion reactor temperatures (≥1000K). Strengthening W with ZrC dispersoids can improve ductility and limit grain growth. However, the effects of ZrC dispersoids on thermomechanical properties and interaction with plasma species are not well understood. We have developed machine-learned interatomic potentials to leverage the predictive capability of first-principles methods for simulations of millions of atoms on the nanosecond scale. We will discuss the development of a Spectral Neighbor Analysis Potential (SNAP) for W-ZrC-H and its performance at fusion reactor temperatures. SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525.
4:40 PM
Effects of Carbide Dispersoids on Helium Bubble Formation in Dispersion-strengthened Tungsten: Xing Wang1; Ashrakat Saefan1; Eric Lang2; Jean Paul Allain1; 1Pennsylvania State University; 2University of Illinois at Urbana-Champaign
Dispersion-strengthened tungsten (DS-W) has shown improved thermomechanical properties compared to W as plasma-facing components (PFC). Exposure to the high fluence of helium (He) is a critical challenge for PFC due to the formation of bubbles and surface nanostructures. Here we investigated the role of carbide dispersoids in bubble formation by in-situ annealing conducted in transmission electron microscopes (TEM). DS-W samples were first irradiated by 2 MeV He at room temperature and only tiny cavities (<1 nm) were found in the material. After annealing at 900 °C for about 30 mins, large bubbles (≥ 3nm) started to appear in both W and carbide. However, few bubbles could be observed at or near W-carbide interfaces, suggesting that the interfaces acted as defect sinks and suppressed bubble formation. Optimizing the composition and distribution of carbide dispersoids can provide a promising approach to designing W-based PFC with superior resistance to He irradiation.
5:00 PM
In-situ Helium Bubble Formation and Thermal Evolution in Lithium Metatitanate: Amy Gandy1; Sam Waters2; Graeme Greaves3; Yiqiang Wang2; 1University of Sheffield; 2UK Atomic Energy Authority; 3University of Huddersfield
Lithium metatitanate (Li2TiO3) is a main candidate for the solid tritium breeder material and will be tested in ITER’s blanket modules. The neutron-induced transmutation of lithium to tritium results in the formation of helium. At blanket operating temperatures, helium may diffuse to, and accumulate at Li2TiO3 grain boundaries, resulting in cracking and loss of structural integrity of the ceramic, and / or trap diffusing tritium, reducing tritium extraction efficiency. To investigate the impact of helium generation in Li2TiO3, we produced suites of samples with different grain sizes and levels of porosity, and irradiated them in-situ in a TEM, up to a fluence of 1e17 He ions/cm2, followed by in-situ thermal annealing up to 1000 °C. In this contribution, we present in-situ TEM images which reveal the impact of grain size and morphology on bubble formation, and the effect of electron irradiation on bubble evolution and transformation to faceted cavities.