Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: Structural Materials Characterization & Modelling II
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Dong Liu, University of Oxford; Peng Xu, Idaho National Laboratory; Simon Middleburgh, Bangor University; Christian Deck, General Atomics; Erofili Kardoulaki, Los Alamos National Laboratory; Robert Ritchie, University of California, Berkeley
Wednesday 8:30 AM
March 2, 2022
Room: 204A
Location: Anaheim Convention Center
Session Chair: Michael Rushton, Bangor University; Conor Oscar Galvin, LANL
8:30 AM Invited
NOW ON-DEMAND ONLY - Development of Direct and Indirect Ab Initio Radiation Damage Models: Par Olsson1; Ebrahim Mansouri1; Qigui Yang1; Elin Toijer1; Par Olsson2; 1KTH Royal Institute of Technology; 2Malmo University
We have developed a full energy range primary damage model that completes the athermal recombination corrected DPA model. All previous indirect models disregard important details of near threshold damage generation, of critical importance for final repository solutions. The model is readily parameterized using ab initio molecular dynamics simulations. Furthermore, we present state of the art direct damage modelling, applicable to both structural and fuel materials. We have developed the first ab initio creation relaxation algorithm in order to access magnetic properties, allow for full chemical complexity and much more. The creation relaxation algorithm essentially mimics cryogenic radiation damage conditions but has great potential in its enabling of assessing the saturation stages that appear in certain finite temperature conditions.Finally, we will discuss how ab initio modelling can be used to directly probe mechanical properties and improve understanding of radiation induced segregation phenomena by coupling stress tests to transport modelling.
9:00 AM
Effective Bias of Cavities in BCC Fe: As Revealed by Atomistic Calculations: Yuhao Wang1; Fei Gao1; Brian Wirth2; 1University of Michigan - Ann Arbor; 2University of Tennessee, Knoxville
A new bias-driven criterion for the critical size of cavity nucleation was recently reported by introducing the cavity bias for interstitials over vacancies. This criterion provides a more universal explanation on cavity behaviors such as incubation period, helium requirement and bimodal size distribution over a much wider temperature range comparing to the classic thermal criterion. The previous study has determined the effective bias for different types of interstitial defects over vacancy defects to voids in BCC Fe and a specific homogenization method was established for larger interstitial clusters due to their one-dimensional diffusion behavior. Considering helium generated from the (n, α) reaction plays an important role in cavity swelling, atomistic calculations were performed to investigate the interaction distance as well as the effective bias for single interstitial, single vacancy and corresponding larger clusters to He bubbles with different sizes and bubble pressure conditions.
9:20 AM Cancelled
Implementation of a Damage Degradation Function for Creep Predictions within a FFT-based Framework: Nathan Beets1; Paul Christodoulou2; M. Arul Kumar1; Ricardo Lebensohn1; Laurent Capolungo1; 1Los Alamos National Laboratory; 2University Of California Sanra Barbara
We present the implementation of a damage degradation function in an elastic-viscoplastic FFT code to capture the effect of porosity on the acceleration of damage during tertiary creep of steels. We obtained an analytical expression for the increased strain-rate due the presence of voids derived from the viscoplastic flow of a voided material with single crystal properties based on an extension of similar analysis for an isotropic viscoplastic matrix. This expression is integrates into the constitutive law for plastic deformation in a full-field elasto-viscoplastic fast Fourier transform (EVPFFT) framework. Creep deformation simulations are performed using the extended implementation for 316H stainless steel. The effects of void nucleation and growth on creep response and rupture lifetime as a function of temperature and applied stress are analyzed. Model predictions are validated against the available experimental observations, and discussed in comparison with other reported damage models.
9:40 AM
NOW ON-DEMAND ONLY – On the Diffusion of Hydrogen Atoms towards Notch Tips in Zirconium Polycrystals: A CPFE Analysis: Alireza Tondro1; Hamidreza Abdolvand1; 1University of Western Ontario
Hydrogen embrittlement is an important degradation mechanism affecting the lifetime of engineering components. The diffusion of hydrogen atoms into the metal lattice can be affected by the localized stresses that develop in the vicinity of service-induced flaws. This study uses a coupled diffusion-crystal plasticity finite element approach to study the redistribution of hydrogen atoms around service-induced flaws in zirconium polycrystals. The effects of texture, grain size, and notch tip geometry on the distribution of hydrogen concentration in the lattice sites are investigated. Results indicate that material texture can significantly affect the distribution of hydrogen atoms as well as the location of maximum hydrogen concentration. It is shown that with changing grain orientations, it is possible to move the location of peak hydrogen concentration away from the notch tip. It is further shown that some triple points and grain boundaries are prone to accommodate higher hydrogen concentration compared to grain interiors.
10:00 AM Break
10:20 AM
Atomistic Simulation Study of the Effect of Hydride Morphology on the Ductility of Polycrystalline Zirconium: Hadi Ghaffarian1; Ye-eun Na1; Dongchan Jang1; 1Korea Advanced Institute of Science and Technology (KAIST)
Zirconium alloys usually undergo degradation in their mechanical properties due to hydride formation during their service time. Although the zirconium hydride is considered a brittle phase, experimental observations indicate the occurrence of plasticity in the embedded hydride within the zirconium matrix, which involves the slip transmission across the Zr/Hydride interfaces. However, the corresponded mechanisms are still unclear due to experimental limitations. Hence, this study aims to look into the deformation mechanism of embedded hydride and clarifies the role of hydride morphology on the ductility of zirconium alloys using the Molecular Dynamics (MD) simulation method. Based on this goal, we constructed nanocrystalline zirconium samples containing hydrides with different sizes, shapes, and orientations and subjected them to the tensile loading. We found that the ductility of the nanocrystalline zirconium samples is governed by the degree of slip transmission across the Zr/Hydride interface, whose transmission ability is altered by the hydride’s spatial configuration.
10:40 AM Cancelled
Oxygen and Carbon Defects in Uranium and Plutonium Nitride: Navaratnarajah Kuganathan1; Conor Galvin1; Robin Grimes1; 1Imperial College London
Actinide mononitrides UN and PuN are attractive nuclear fuels for future reactor systems due to their relatively high thermal conductivities. However, impurities are known to impact this and other properties. Using atomic scale computer simulation, based on density functional theory, we investigate the structures, defect energetics and trapping of C and O impurities in UN and PuN. Gas phase, oxide and carbide reference states are considered as well as Schottky and Frenkel disorder processes, magnetic order and non-stoichiometry. The incorporation of O and C at interstitial and lattice sites is predicted along with di-vacancy and a variety of dumbbell configurations. While solution from the oxide and carbide is, of course, unfavourable, there are clear differences between energies in UN and PuN. These are demonstrated though the generation of defect concentrations as a function of temperature and partial pressures.
11:05 AM
Temperature Sensitive Dislocation Dynamics Modeling of Hardening and Embrittlement: Aaron Kohnert1; Laurent Capolungo1; 1Los Alamos National Laboratory
Irradiation induced shifts in the nil ductility temperature can lead to significant loss of toughness and brittle failure for nuclear reactor components at relatively low exposure levels. In addition to embrittlement concerns in pressure vessel steels for existing reactors, this may complicate the use of similar materials such as ferritic steels in future reactor concepts. In this study, we investigate the deformation behavior of iron using an advanced, 3d dislocation dynamics model leveraging realistic dislocation networks comprised of edge, screw, mixed character, and junction segments. The model includes a temperature dependent, anisotropic Peierls barrier for screw dislocations paired with thermally activated mechanisms including both cross-slip and climb. Dislocation microstructures are loaded in uniaxial tension and compared to systems which include varying densities of radiation induced precipitates. Application across many loading conditions provides direct quantification of shifts in the yield point as a function of temperature, strain rate, and precipitate content.
11:25 AM
Radiation-Induced Segregation in Binary Alloy Systems Examined Via Phase Field Simulations: Daniel Vizoso1; Chaitanya Deo1; Remi Dingreville2; 1Georgia Institute of Technology; 2Sandia National Laboratories
Materials exposed to a radiation environment experience microstructural changes due to the local atomic shuffling and defect formation. One evolutionary mechanism occurring in such environment in metallic alloys is radiation-induced segregation (RIS), where atomic mixing and radiation-enhanced diffusion can disrupt the random order of a miscible alloy and induce segregation or depletion of solutes. These processes occur over a wide range of time- and length-scales, which makes modeling RIS via conventional atomistic methods computationally challenging. In this presentation, we discuss a mesoscale phase-field model for RIS in binary alloys motivated by atomistic results that includes the consideration of irradiation conditions (electron vs. ion irradiation; flux), dislocations as sinks, and the formation and growth of defect clusters. We illustrate our results on an exemplar binary alloy and discuss their dependence on dose rate, defect sinks, and the damage mechanism. SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525.