Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: On-Demand Oral Presentations
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
Monday 8:00 AM
March 14, 2022
Room: Nuclear Materials
Location: On-Demand Room
A Comprehensive Study of Responses of SiC Materials to Neutron Irradiation for ATF Cladding Application: Takaaki Koyanagi1; Christian Petrie1; Jose' Arregui Mena1; Hsin Wang1; Yutai Katoh1; 1Oak Ridge National Laboratory
Development of SiC-based composites continues for use in accident-tolerant fuel cladding for light water reactors (LWRs) because of its inherent advantages of low neutron absorption, retention of strength following irradiation, and high-temperature capability. Although responses of SiC materials to neutron irradiation have been extensively studied, there remains a knowledge gap in predicting the performance of SiC cladding under irradiation. This paper presents recent findings from evaluations of SiC materials irradiated under LWR-relevant temperature and dose conditions. The research subjects include (1) high-dose irradiation effects, (2) irradiation-induced stress and dimensional changes under temperature and neutron flux gradients, and (3) stored energy release during post-irradiation annealing. These studies reach a common conclusion that mechanistic understanding and modeling of irradiation-induced lattice swelling is important to assessing the performance of SiC-based cladding. This study was supported by US Department of Energy–Nuclear Energy and Westinghouse Electric Company/General Atomics FOA projects.
Advances in SiGAŽ Development for Nuclear Applications: Sean Gonderman1; Kirill Shapovalov1; George Jacobsen1; Lucas Borowski1; Shaobin Fan1; Rolf Haefelfinger1; Christian Deck1; Jack Gazza1; Christina Back1; 1General Atomics
General Atomics–Electromagnetics (GA-EMS) is developing SiGAŽ composite technology, an engineered SiC fiber reinforced, SiC matrix (SiC-SiC) composite, for light water reactor applications to deliver improved normal operational performance and enhanced safety. Progress will be reported on the scaling of SiGAŽ cladding fabrication for lead test rods (LTRs) by 2025, including increases in cladding length capacity and throughput through different fabrication steps. Design work on advanced SiGAŽ fuel rods to support high burn-up, high enrichment fuel with SiC-SiC-based cladding show encouraging results, improved thermo-mechanical properties, meeting of dimensional specifications at longer lengths, and promising hydro thermal corrosion performance. These advances are preparing SiGAŽ cladding for upcoming irradiations at Southern Companies Vogtle plant and Idaho National Lab’s Advanced Test Reactor. In addition to cladding development, SiGAŽ structures are being brought to bear on other in core structures, further advancing the concept of accident tolerant fuel towards a fully accident tolerant core.
Creep Performance of IronClad Accident Tolerant Fuel Cladding: Evan Dolley1; Wanming Zhang1; Dan Lutz1; Russ Fawcett1; Raul Rebak1; 1GE Research/GE Power
The nuclear power industry has been using zircaloy alloy clad fuel for over 60 years. Since the Fukushima accident, the US Department of Energy has been working with General Electric (GE) to develop accident tolerant fuels to make the current fleet of reactors safer to operate. GE is evaluating IronClad or FeCrAl alloy as cladding for UO2 pellets. Since FeCrAl ferritic alloys have not previously been used in light water reactors, it is important to characterize their behavior over the entire fuel cycle. GE has a test method for creep testing of IronClad tubing as a function of temperature and hoop stress. Results show that powder metallurgy IronClad tubing showed diametral creep rates at temperatures up to 800°C that are several times lower than those of cladding tubes fabricated from Zircaloy-2. Creep rates of IronClad tubing have also been shown to be dependent on material chemistry and resultant microstructural conditions.
Effect of Free Surface on Displacement-cascade Damage in Neutron Irradiated Nickel: Michele Fullarton1; Giridhar Nandipati1; Ram Devanathan1; David Senor1; 1Pacific Northwest National Laboratory
During reactor operation, the production of point defects from displacement cascades due to energetic neutron-atom collisions and their accumulation rapidly degrades the physical and mechanical properties of Nickel (Ni) and its alloys. In this study, we explored the differences in the cascade dynamics in the bulk and the near-surface region in pure Ni using molecular dynamics simulations, as a function of temperature (300, 425 and 525 K) and primary-knock-on atom (PKA) energy (1 to 10 keV). Additionally, the depth and direction of PKAs in the near-surface region were studied. In all cases, PKA direction significantly impacted the count and size distribution of surviving defect clusters. We will present a detailed analysis of cascade dynamics and point defect production to illustrate the influence of a free surface.
Effects of Low-temperature Neutron Irradiation, Hydrogen Charging, and Post-weld Heat Treatment on Tensile Properties of Welded Zircaloy-4: John Echols1; Nate Reid1; Lauren Garrison1; 1Oak Ridge National Laboratory
Zircaloy-4 enjoys excellent corrosion, radiation, and mechanical properties. Use has historically been limited mostly to fuel cladding above 300°C. Expanding the usage of zircaloy-4 has been proposed for pressure vessels, test reactors, and reprocessing plants. However, hydrogen, welding, and neutron irradiation effects on microstructure and mechanical properties need to be qualified and deconvoluted for these new operating regimes. This work investigates zircaloy-4 with/without TIG welding, post-weld heat treatment (PWHT), and hydrogen charging - neutron irradiated in the High Flux Isotope Reactor (target doses 1021 and 1022 n/cm2) at low temperature (60-90°C) at Oak Ridge National Laboratory. Room-temperature tensile and hardness testing was performed. Emphasis is placed on understanding ductility as a function of weld, PWHT, hydrogen content, and dose. The importance of PWHT, H-charging, irradiation, and welding and their effects on microstructure (including formation of the undesirable “blocky alpha” phase) and mechanical behavior will be discussed in detail.
Uranium Nitride/Uranium Boride Composite Materials: Joel Turner1; James Buckley1; Robert Worth1; Tim Abram1; 1University of Manchester
Uranium nitride fuel is of interest as an alternative to uranium dioxide due to the higher uranium density and thermal conductivity it offers. The primary challenge to its use in light water reactors is the relatively low temperature at which UN undergoes hydrolysis in steam. We will outline work in which we have increased this onset temperature by over 100C via the addition of a uranium diboride phase. The addition of UB2 has further potential benefits, as it has a very high thermal conductivity, operates as a burnable absorber while not removing significant uranium from the UN matrix. Manufacture and initial characterisation data are presented, alongside steam reaction measurements and post-oxidation characterisation.
Simulation of Shearing-induced Edge and Interfacial Fractures in U-10Mo Monolithic Fuel Plates: Lei Li1; Kyoo Sil Choi1; Kenneth Johnson1; Vineet Joshi1; Ayoub Soulami1; 1Pacific Northwest National Laboratory
Low-enriched uranium metal alloyed with 10wt% molybdenum (U-10Mo) is being developed and qualified by the NNSA to replace high-enriched uranium fuel due to its ability to meet the neutron flux requirements in U.S. high-performance research reactors. Shearing is a critical step in the U-10Mo monolithic fuel fabrication, which trims off redundant interfacially bonded materials to meet the target dimension. This work proposes a finite element model to predict the edge and interfacial fractures induced by the shearing process. The proposed model is demonstrated and validated on a HIP-bonded U-10Mo monolithic fuel plate, where the bonding strength and energy release rate of aluminum/aluminum interface for the cohesive model are determined using peel and shear testing data. A sensitivity study various shearing conditions on the fracture characteristic was conducted. Based on the simulation results, the optimal shearing parameters were determined and suggested to improve the interfacial fracture resistance during the shearing process.
Phase Stability in FeCrAl Alloys: Mapping the Miscibility Gap and Understanding the Impact of Alpha Prime Precipitation on Material Properties: Rajnikant Umretiya1; Andrew Hoffman1; Raul Rebak1; 1GE Research
FeCrAl alloys are a leading candidate class of alloys for accident tolerant fuel cladding due to their good mechanical properties and hydrothermal corrosion resistance coupled with their excellent high temperature steam oxidation resistance. One of the largest unknowns in FeCrAl alloys is understanding the phase stability of this system at temperatures below the ~500°C miscibility gap. In this presentation we will discuss work that shows the impact of chemistry on thermodynamics and kinetics, experimental mapping of the FeCrAl phase diagrams, and how α’ precipitation impacts material properties including corrosion and mechanical properties.
Microstructure and Surface Chemistry of FeCrAl Alloys Accident Tolerant Fuel Cladding Subjected to Fast Heating Rate in Aqueous Environment: Rajnikant Umretiya1; Donghwi Lee2; Mark Anderson2; Raul Rebak1; Jessika Rojas3; 1GE Research; 2Universty of Wisconsin-Madison; 3Virginia Commonwealth University
FeCrAl alloys, such as APMT and C26M, are excellent cladding candidates for accident-tolerant fuel (ATF) systems in light water reactors due to their high resistance to oxidation as a result of the formation of a protective Cr2O3 and Al2O3 scale at high temperatures. In order to understand their evolution at the early stage of high-temperature exposure in short periods, samples were exposed to flow boiling testing and heated to ~400 ēC within a second. Using scanning electron microscopy (SEM), ~100-200 nm thick oxide layer was confirmed. The surface chemistry of both alloys was characterized layer by layer using X-ray photoelectron spectroscopy (XPS) depth profiling. The results indicated a layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer. After the tests, the average micro-hardness values were increased, and that was confirmed with higher mechanical properties acquired by ring compression testing.
Interface Characterization of an Explosion Welded Stainless Steel-clad Plate for Neutron Irradiation Studies: Nathan Reid1; John Echols2; Lauren Garrison2; Jean Paul Allain3; 1University of Illinois at Urbana-Champaign; 2Oak Ridge National Laboratory; 3Pennsylvania State University
Explosion welding can create a clean weld between a thin, corrosion resistant layer of stainless steel and carbon steel. This presents a high-throughput, rapid avenue for generating composite materials for nuclear pressure vessels and fusion structural components. The joint interface is complex and distinct from liquid-phase and other bonding techniques. Therefore, this joint has been rigorously qualified with mechanical testing following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The interface is examined by microscopy, Charpy testing, patterned microhardness testing, uniaxial tensile testing, and compressive shear testing before irradiation. Compressive shear testing is used to determine notch placement in tensile specimens of the bi-layer material for tensile lap-shear specimens. Bi-layer tensile specimens were neutron irradiated and uniaxial tensile tested.
Diffusion in Doped and Undoped Amorphous Zirconia: Megan Owen1; Michael Rushton1; Lee Evitts1; Antoine Claisse2; Mattias Puide2; William Lee1; Simon Middleburgh1; 1Bangor University; 2Westinghouse Electric Sweden AB
Grain boundaries in ZrO₂ formed on Zr-based nuclear fuel cladding may be high diffusion pathways for corrosion species such as oxygen and hydrogen. Alloying element segregation has been observed experimentally towards the grain boundaries in ZrO₂, facilitating the development of amorphous regions, with faster diffusivity than the bulk. Undoped and doped amorphous ZrO₂ have been simulated successfully using molecular dynamics, to analyse structures formed, alongside the diffusivity of oxygen as a function of dopant radius and concentration. Comparisons with tetragonal undoped and doped zirconia were made throughout, highlighting differences in diffusion based on the structure of the system [1]. Further work will involve analysing the role of dopants on the electrochemical behaviour, potentially stabilising the amorphous systems.[1] M. W. Owen et al., “Diffusion in doped and undoped amorphous zirconia,” J. Nucl. Mater., 2021, doi: 10.1016/j.jnucmat.2021.153108.
Effects of Steel Composition and Grain Size on Diffusion with Neodymium: Brian Bettes1; Yi Xie1; 1Purdue University
Investigations into the nuclear fuel-cladding chemical interactions between neodymium and four steels were conducted via diffusion couple experiments. The effects of steel composition and grain size were of primary interest in this investigation. Neodymium was chosen for these tests because it constitutes the highest concentration of lanthanide fission products present in irradiated metallic fuel. The four steels were conventional type 316 stainless steel (C316), advanced manufactured steel (A316) that is in the same composition as C316 but a larger grain size, iron-chromium (Fe-Cr) based oxide dispersion-strengthened (ODS) steel, and iron-chromium-nickel (Fe-Cr-Ni) based ODS steel. Variation in grain size between A316 and C316 was found to have significant effects on the diffusion of Fe and Ni from the alloys. Diffusion mechanisms were found to be enhanced by the presence of Ni, which was highly reactive with Nd.
Development of PVD Cr Coatings for Hydrothermal Corrosion Resistance of SiC-SiCf Fuel Cladding in LWRs: Kyle Quillin1; Hwasung Yeom1; Tyler Dabney1; Evan Willing1; Taeho Kim1; Sergey Chemerisov2; Christian Deck3; Adrien Couet1; Kumar Sridharan1; 1University of Wisconsin-Madison; 2Argonne National Laboratory; 3General Atomics
SiC-SiCf composite is being considered as a next-generation fuel cladding material for light water reactors (LWRs) based on its superior mechanical strength and oxidation resistance at high temperatures experienced during a potential loss-of-coolant accident (LOCA) compared to currently used Zr-alloys. However, SiC is prone to hydrothermal corrosion under normal LWR operating conditions. We are investigating the deposition of Cr coatings on SiC to address this concern. The coatings are being developed using a variety of state-of-the-art physical vapor deposition techniques. Through detailed microstructural analysis, we explain the effect of deposition parameters on coating microstructure and density. Nanoindentation and micropillar compression tests have been conducted to evaluate the mechanical properties of the coatings, and prototypical autoclave tests have been performed to understand their corrosion performance. The effects of water radiolysis on corrosion will also be discussed.
Role of Powder Microstructure and Mechanical Properties on Deposition and Properties of Cold Spray Cr Coatings: Tyler Dabney1; Kyle Quillin1; Hwasung Yeom1; Kumar Sridharan1; 1University of Wisconsin-Madison
Thin Cr coatings are being developed as oxidation-resistant barriers for zirconium-alloy fuel rods to improve safety tolerance for the current Light Water Reactor (LWR) fleet. One way to deposit these coatings is through cold spray, a solid-state deposition technique in which micron-sized particles are accelerated at supersonic velocities towards a substrate surface using a high-pressure gas. Upon impact, the particles plastically deform at high strain rates to form a coating. The microstructure of the coating is strongly dependent on the microstructure and mechanical properties of the feedstock Cr powder. We will present characterization results of Cr feedstock powders manufactured by various methods and consequently with different mechanical properties. The mechanical behavior of Cr powder particles and the resultant coatings were evaluated by small-scale mechanical testing. Finally, the structure-property relationship will be discussed with insights of how initial powder microstructure affects the deposition and mechanical behavior of cold spray Cr coatings.
Manufacturing of Oxide Dispersion Strengthened (ODS) Steel Fuel Cladding Tubes Using Cold Spray Technology: Hwasung Yeom1; Vishnu Ramasawmy1; Xinwu Liu1; David Hoelzer2; Stuart Maloy3; Peter Hosemann4; Kumar Sridharan1; 1University of Wisconsin Madison; 2Oak Ridge National Laboratory; 3Los Alamos National Laboratory; 4University of California-Berkeley
Solid-state, powder-based cold spray manufacturing approaches to fabricate oxide dispersion strengthened (ODS) steel fuel cladding tubes for advanced nuclear designs have been investigated as an alternative of the conventional multi-step powder extrusion and annealing process. The study includes extensive parametric investigation of cold spray deposition process, powder size distribution, as well as post-deposition heat treatments to achieve high-density ODS steel. Engineered feedstock powders such as gas-atomized powder, ball-milled powder, cryogenically milled powder, and thermally treated powder were utilized in the manufacturing routes to design microstructurally optimized ODS steel. The feasibility of cold spray deposition of functional coatings (e.g., corrosion and wear-resistant) on these ODS cladding tubes for the harsh environments of a nuclear reactor will be presented. Finally, data for benchmarking properties of ODS steel materials produced by cold spray process with those manufactured using other routes will be discussed.
Progress on Experimental Investigation of Degradation Mechanisms of ATF Coated Cladding under Transient Conditions: Hwasung Yeom1; Tyler Dabney1; David Kamerman2; Michelle Bales3; Logan Crevelt4; Zhen Li4; Anthony Evans4; Brent Heuser4; Kumar Sridharan1; 1University of Wisconsin Madison; 2Idaho National Laboratory; 3U.S. Nuclear Regulatory Commission; 4University of Illinois Urbana-Champaign
Chromium coatings on Zr-alloy cladding are being considered for implementation by nuclear industries to enhance reactor safety during accidents as well as to achieve extended burn-up with improved fuel cycle economics. However, there are still knowledge gaps for robust data sets to support licensing and qualification of the designs, including fuel degradation under design-basis accidents. This research project aims at investigating the thermal, mechanical, and irradiation response of Cr-coated Zr-alloy cladding under prototypical reactivity-initiated accidents (RIA), in comparison to that of uncoated Zr-alloy cladding. This presentation will introduce recent progress of this program including experimental plan at TREAT at INL, fabrication of Cr-coated Zr-alloy tubes using physical vapor deposition and cold spray deposition methods, design of static-water test vehicles, and preliminary test results. The experimental data on response of fueled coated cladding under RIAs will provide validation of results of modeling activities for the ATF fuel designs during reactor transients.
Microstructure, Mechanical Properties, and Irradiation Response of Fe-Cr-Ni-based Multi-principal Element Alloys: Marcus Parry1; Cheng Sun1; Wen Jiang1; Boopathy Kombaiah1; Colin Judge1; Seongtae Kwon1; Ovidiu Toader2; Gary Was2; Taylor Sparks3; 1Idaho National Laboratory; 2University of Michigan; 3University of Utah
The implementation of advanced nuclear design concepts requires advanced structural materials to withstand the associated harsh conditions, including elevated temperatures, pressures, and radiation doses, while some designs also involve new coolant environments. Multi-principal element alloys (MPEAs) may address these challenges as they have demonstrated extraordinary strength, hardness, oxidation and corrosion resistance, thermal stability, and radiation tolerance, driving appeal as potential reactor structural materials. However, composition-microstructure-property relationships among MPEAs remain unclear, particularly relating to irradiation damage. In this study, four spark plasma sintered MPEAs in the Al-Cr-Fe-Ni-(Cu,Mn) system are investigated to clarify these relationships. At Michigan Ion Beam Laboratory, accelerator technologies were utilized to irradiate the alloys with 2.0 MeV protons at 400 ℃ to damage levels reaching 5 displacements-per-atom. Phase stability and nanomechanical response of pre- and post-irradiated specimens are examined to elucidate correlations among alloy chemistry, microstructure, and irradiation damage tolerance.
Deconvoluting Properties of Additively Manufactured Alloy 718 Utilizing Coupled Microscopy and Machine Learning: Stephen Taller1; Ty Austin2; Kurt Terrani1; 1Oak Ridge National Laboratory; 2University of Tennessee - Knoxville
Ni-based superalloys are a primary candidate alloy class for high temperature applications because of their intrinsic resistance to creep, their adequate resistance to corrosion and their high strength. The poor machinability and extensive work hardening make additive manufacturing (AM) an attractive option to produce geometrically complex components with distinct microstructures from wrought alloys. The microstructure of an AM-produced superalloy 718 was characterized for precipitates, pores, and dislocations using a dynamic segmentation convolutional neural network (DSCNN) with pixel-wise defect-detection algorithms. Solute segregation contributed to multiple unwanted phases, e.g. Laves, in the as-built microstructure. Three heat treatments were performed to anneal unwanted precipitates, metastable precipitates, or solution anneal the microstructure. Uniaxial tensile tests were performed at room temperature and 300-600°C. The combination of heat treatments and rapid characterization of the microstructure using the DSCNN improve the understanding of the relative contribution of each phase to the strength of AM superalloy 718.
The Interaction between an Extended Edge Dislocation and a Helium Bubble in Copper: Wu-Rong Jian1; Shuozhi Xu1; Yanqing Su2; Irene Beyerlein1; 1University of California, Santa Barbara; 2Utah State University
Exposed to irradiation environment, structural metals in nuclear reactors suffer from the bombardment of high-energy particles, generating large amounts of nano-sized He bubbles. Such bubbles can interact with the gliding dislocations and then give rise to the strain hardening of metals. By performing molecular dynamics simulation in Cu, we investigate the interaction between an edge dislocation and He bubble. Instead of the conventional bypass modes over a He bubble, i.e., bubble cutting and dislocation climb, a new multi-step-bypass (MSB) maneuver is found at room temperature when the He-vacancy ratio is large enough and the ratio of the bubble spacing to its diameter is reduced below a threshold. For MSB, the entire dislocation changes its glide plane to overcome the bubble, which is promoted by higher temperature and larger He-vacancy ratio in the bubble. Compared to the conventional bypass mechanisms, MSB is more energetically favorable.
Phase-field Simulations of Fission Gas Bubbles in High Burnup UO2 to Inform Engineering-scale Fuel Performance Modeling: Larry Aagesen1; David Andersson2; Sudipta Biswas1; Michael Cooper2; Kyle Gamble1; Wen Jiang1; 1Idaho National Laboratory; 2Los Alamos National Laboratory
To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized that pulverization is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted, and the results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.
Unraveling the Early Stage Ordering of Krypton Solid Bubbles in Molybdenum: A Multi-modal Study: Ericmoore Jossou1; Anton Schneider2; Cheng Sun3; Yongfeng Zhang2; Shirish Chodankar1; Dmytro Nykypanchuk1; Jian Gan3; Lynne Ecker1; Simerjeet Gill1; 1Brookhaven National Laboratory; 2University of Wisconsin; 3Idaho National Laboratory
Self-organization of defects, such as fission gas bubbles in nuclear fuel alloys can lead to higher capacity for fission gas storage and help mitigate swelling caused by fission gases in nuclear reactors. We report the physical mechanism of self-organization of Kr in Mo under ion implantation. The ion fluence and temperature-dependent formation of Kr solid bubble superlattice (SBS) in Mo was investigated using synchrotron-based small-angle x-ray scattering and transmission electron microscopy in combination with atomic kinetic Monte Carlo (AKMC) modeling. Early-stage self-organization of solid bubbles as a function of dose and temperature will be discussed. The mechanism of Kr SBS is compared with He bubble ordering in Mo matrix. AKMC simulations is used to correlates the binding energy of Kr with the mobility and the eventual stability of Kr SBS. Overall, our work sheds light on the formation mechanism of noble gas superlattice towards the development of radiation tolerant materials.
Cancelled
Modeling High-temperature Corrosion of Zirconium Alloys Using the Extended Finite Element Method: Wen Jiang1; Louis Bailly-Salins2; Benjamin Spencer1; Adrien Couet2; 1Idaho National Laboratory; 2University of Wisconsin-Madison
The oxidation process of zirconium-alloy cladding fundamental in LWR fuel performance. The resultant oxidation layer can affect the cladding’s thermal and mechanical properties. In current nuclear fuel performance codes, oxidation modeling is limited by the lack of coupling with mechanics, thus impeding a proper description of high-temperature oxidation’s impact on mechanical properties. A recent development in the extended finite element method (X-FEM) in the nuclear fuel performance code BISON enables precision tracking of interfaces at relatively low computation cost. The Coupled-Current Charge Compensation model, a physically-based zirconium alloy corrosion model, was implemented in BISON to predict oxide layer growth and oxygen concentration profiles. The oxygen accumulation leads to reduced ductile thickness of the cladding, and eventually to cladding failure. The ability to couple X-FEM with the mechanics in BISON is a first-of-a-kind achievement in oxidation modeling and would enable more comprehensive simulations of fuel rod behavior during LOCAs.