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DOE OSTI2025

Spatial mapping of the localized corrosion behavior of a magnesium alloy AZ31B tungsten inert gas weld

Sections of a magnesium alloy, AZ31B, joined with tungsten inert gas (TIG) welding, were examined with scanning electrochemical microscopy (SECM) and scanning Kelvin probe force microscopy (SKPFM) to investigate corrosion mechanisms by correlating observed corrosion behavior with weld-affected microstructural variations. Insight into the changing nature of the galvanic couples between weld zones and at localized microgalvanic sites were investigated using SECM and SKPFM to map both electrochemically active regions and Volta potential differences across the weld-affected zones. The formation of an Al-Zn solidification network in the fusion zone (FZ) at and near the TIG weld epicenter differs from the outer heat-affected zone (HAZ), where intermetallic particles (IMPs) are the notable secondary phase from the magnesium matrix. These microstructures were mapped with SKPFM before and after brief exposure to a salt solution, revealing micro-galvanic couples as the main driving force to corrosion initiation and propagation within each zone. The IMPs and Al-Zn solidification network act as strong cathodes and govern the corrosion processes. The galvanic coupling and evolution of the intrinsic corrosion behavior between the weld zones is explained by monitoring the hydrogen evolution reaction (HER) with SECM over time. Anodically induced cathodic activation is confirmed for this welded material, as micro-galvanic couples between microstructural features are found to transition over time to broad electrochemically active areas within the weld-affected zones, resulting in polarity reversal as time of exposure proceeds.

Anodically induced cathodic activation↗

DOE OSTI2021

Corrosion behavior in aluminum/galvanized steel resistance spot welds and self-piercing riveting joints in salt spray environment

Here, this paper underlined the corrosion behavior of aluminum alloy/galvanized steel joints fabricated by resistance spot welding (RSW) and self-piercing riveting (SPR) exposed to different salt-spray cycles. It was found that the crevice generated by different joining methods has an important impact on corrosion behavior. Compared with the RSW joint, less corrosion happened on the coupled regions in SPR joints because of the higher local pH value resulted from the smaller crevice. The crevice, however, has less impact on the types of corrosion products. Galvanic corrosion happened in the coupled region for both RSW and SPR joints. The Pourbaix diagram explained the stable corrosion products of Al and Zn—Fe alloy, which are Al 2 O 3 , ZnO, and Fe 3 O 4 . Additional corrosion products observed from the experiment included α-Al(OH) 3 , Ca 4 Al 2 (CO 3 )(OH) 12 (H 2 O) 5 , Zn 5 (OH) 8 Cl 2 (H 2 O), Zn 5 (CO 3 ) 2 (OH) 6 , and Fe 2 (OH) 2 (CO 3 ), which can be understood by considering the metastable corrosion products, Pourbaix diagram, and possible chemical reactions.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Deformation behavior of metastable β-type Ti-25Nb-2Mo-4Sn alloy for biomedical applications

The deformation behavior of metastable β-type Ti–25Nb–2Mo–4Sn (wt%) alloy subjected to different thermo-mechanical treatments was discussed by the combining results from transmission electron microscope, tensile test and in-situ synchrotron X-ray diffraction. Visible “double yielding” behavior, which is characterized by the presence of stress-plateau, was observed in the solution treated specimen. Upon a cold rolling treatment, the Ti–25Nb–2Mo–4Sn alloy performs nonlinear deformation because of the combined effects of elastic deformation and stress-induced α" martensitic transformation. After the subsequent annealing, the β phase is completely stabilized and no stress-induced martensitic transformation takes place on loading due to the inhibitory effect of grain boundaries and dislocations on martensitic transformation. As a result, the annealed specimen exhibits linear elastic deformation.

Guo, S.↗

DOE OSTI2025

Characterization of the fatigue threshold behavior of UHMWPE

Ultra-high-molecular-weight-polyethylene (UHMWPE) has been the material of choice for bearings in total joint replacements (TJRs) for decades as a result of its excellent wear resistance, chemical inertness, energetic toughness, low friction, and biocompatibility. Utilization of this polymer in orthopedic devices requires oxidation, wear, and fatigue resistance. Balancing these important properties by tailoring processing techniques and modulating microstructural features has been an ongoing endeavor in the field. Research into the clinical applications of UHMWPE has primarily focused on the challenges of wear and oxidation while studies into the realm of fatigue have been more limited. Literature gaps exist in fully understanding the fatigue crack initiation near notches or propagation of small existing flaws in UHMWPE used in TJRs. In particular, the characterization of the fatigue thresholds and near-threshold fatigue behavior of orthopedic grade UHMWPE has yet to be thoroughly explored. In this work, we characterized the fatigue crack arrest threshold of clinically-relevant UHMWPE formulations. Correlations between the fatigue thresholds and bulk mechanical properties as well as microstructural properties were examined across these medical resins. The important role played by crosslinking in influencing the fatigue performance of UHMWPE is highlighted in this study. In addition, it is established that J-integral fracture toughness is the best predictor of fatigue thresholds and could possibly be used as a stand-in metric for fatigue performance if thresholds cannot be directly ascertained. Finally, this study corroborates that the true constitutive parameters best describe the mechanical behavior of UHMWPE.

Cross-linking↗

DOE OSTI2025

Dislocation nano-hydrides in nickel: Nucleation, evolution and effects on dislocation behaviors

Nano-hydrides have been predicted to precipitate at the core of edge dislocations in the Ni-H system, a mechanism that may promote hydrogen embrittlement. However, nano-hydride nucleation, growth, and effects on dislocation behavior have seldom been explored. This work combines molecular dynamics grand canonical Monte Carlo (MD-GCMC) simulations and continuum modeling to uncover a wide range of phenomena linked to dislocation nano-hydrides. Simulations reveal that nano-hydrides can be stabilized at dislocation cores with all character angles, including screw segments, due to the hydrostatic stresses around the cores of the Shockley partials. Nano-hydride nucleation takes place in these regions, and growth is dictated by the character angle θ of the perfect dislocation. The equilibrium stacking fault width d eq varies dynamically to increase the local hydrostatic stress field and facilitate the formation of the nano-hydride, forming a constriction-like feature and leading to three distinct behaviors: d eq decreases for θ > 30°, d eq remains unchanged for θ = 30°, and d eq increases for θ < 30°. Remote hydrostatic and Escaig stresses are also shown to influence the nucleation stage, implying stress concentrations such as those ahead of crack tips may facilitate nano-hydride precipitation. Moreover, we identify a new hydrogen-induced 60° dislocation reaction that emits a Shockley partial on a conjugate plane, with potential implications for twin nucleation. Testable predictions from this study are then used to reinterpret previous results from the literature. In conclusion, these findings provide a comprehensive framework to assess nano-hydride formation and evolution at dislocations in nickel and other face-centered cubic metals, with important implications to hydrogen embrittlement.

Dislocations↗

DOE OSTI2026

Topology-driven compressive behavior of Inconel 718 lattice structures with Z-strut reinforcement fabricated by laser powder bed fusion

This study investigates the compressive deformation behavior and mechanical performance of Inconel 718 lattice structures fabricated by laser powder bed fusion (LPBF). Four unit-cell topologies—BCC, FCC, BCCZ, and FCCZ—were designed with a unit-cell size of 3 mm and fabricated under identical process conditions to isolate the effect of topology. Measured relative densities ranged from 14.65% to 17.72%. Compressive testing showed that Z-strut-reinforced topologies (BCCZ: 54.6 MPa; FCCZ: 80.2 MPa) exhibited higher strength than their unreinforced counterparts, which may be associated with mixed-mode deformation behavior enabled by the vertically aligned Z-struts. Finite element simulations and Digital Image Correlation (DIC) analysis support the observation of a transition from node-dominated deformation in BCC/FCC to mixed-mode deformation in BCCZ/FCCZ. These findings suggest that unit-cell topology is a key design variable for tailoring deformation mechanisms in LPBF lattice structures.

36 MATERIALS SCIENCE↗

DOE OSTI2017

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

U 3 Si 2 behavior in H 2 O environments: Part II, pressurized water with controlled redox chemistry

Recent interest in U 3 Si 2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H 2 O environments is sparse in available literature. The behavior of U 3 Si 2 in H 2 O containing atmospheres is investigated and presented in a two-part series of articles. This work examines the behavior of U 3 Si 2 following exposure to pressurized H 2 O at temperatures from 300 to 350 °C. Testing was performed using two autoclave configurations and multiple redox conditions. Use of solid state buffers to attain a controlled water chemistry is also presented as a means to test actinide-bearing systems. Buffers were used to vary the hydrogen concentration between 1 and 30 parts per million H 2 . Testing included UN, U 3 Si 5 , and UO 2 . Both UN and U 3 Si 5 were found to rapidly pulverize in less than 5 h at 300 °C. Uranium dioxide was included as a control for the autoclave system, and was found to be minimally impacted by exposure to pressurized water at the conditions tested for extended time periods. Testing of U 3 Si 2 at 300 °C found reasonable stability through 30 days in 1–5 ppm H 2 . However, pulverization was observed following 35 days. The redox condition of testing strongly affected pulverization. Characterization of the resulting microstructures suggests that the mechanism responsible for pulverization under more strongly reducing conditions differs from that previously identified. Hydride formation is hypothesized to drive this transition. In conclusion, testing performed at 350 °C resulted in rapid pulverization of U 3 Si 2 in under 50 h.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Analysis and modeling of the equilibrium behaviors of U and Pu in molten LiCl-KCl/Cd system at 500 °C

In the field of pyroprocessing of used nuclear fuels, liquid cadmium is applied in different ways to recover transuranics from molten salt, while effecting some degree of separations between transuranics and lanthanides. In this work, a model of the separation performance between uranium and plutonium across the phase boundary between molten salt and liquid cadmium is presented. The salt is the LiCl-KCl eutectic with solute concentrations of UCl 3 and PuCl 3 . The cadmium is the liquid cadmium phase field from the Cd-U-Pu ternary alloy system. Lastly, the model used to describe the behavior of uranium and plutonium in the salt/cadmium system is applied to describe the behavior of plutonium and lanthanide in the salt/cadmium system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications

Incoloy 800HT and Inconel 617 have been selected as candidate structural alloys for the hightemperature gas-cooled reactor (HTGR) concept. Helium, the primary coolant, contains impurities (e.g., H2O and CH4) that can induce corrosion reactions at high temperatures, which in turn can affect the tribological behavior of components in sliding contact such as valves and control-rod drive systems. This paper presents results from the study of the tribological behavior of both alloys before and after conditioning them in either an impure-helium oxidizing environment or in an air environment. Both alloys were conditioned for 22 days at elevated temperatures in a once-through helium loop with 4 ppmv H2O and tested subsequently at elevated temperatures with a pin-on-disk tribometer in an air environment 650 degrees C and 750 degrees C for 800HT; 850 degrees C and 900 degrees C for 617 - with various applied loads - 1 N, 2 N and 5 N. SEM-EDS analysis revealed that conditioning the samples in an oxidizing environment leads to the formation of a mixed Fe/Cr-oxide on alloy 800HT and a Cr-oxide on alloy 617, both increasing the wear resistance compared to that of as-received samples. Alloy 617 exhibited lower steady-state friction coefficients compared to those of alloy 800HT. There was a significant decrease in the scatter of the steadystate friction coefficient of the conditioned samples compared to that of the unconditioned samples. The steady-state friction coefficient for alloy 800HT and 617 were found to be 0.53 +/- 0.07 and 0.25 +/- 0.04, respectively. The wear resistance of alloy 800HT is approximately one order of magnitude lower than that of alloy 617 in all cases that exhibited measurable wear. After sample conditioning, the wear volumes measured at low loads were undistinguishable from the unworn background, a result attributed to the formation of a compacted glaze layer under high temperatures and high contact stresses.

Alloys 800 and 617↗

DOE OSTI2019

Damage mechanics modeling of the non-linear behavior of SiC/SiC ceramic matrix composite fuel cladding

A damage model is developed and validated with experimental data for the non-linear mechanical behavior of SiC/SiC composite materials in nuclear applications. The model extends a continuum damage mechanics formalism, where explicit property degradation and volumetric swelling as a result of neutron irradiation are additionally included. The model is shown to be validated with out-of-pile experiments, and then applied to off-normal and high-burnup operating conditions associated with loss of cooling and extended radiation exposure, respectively. Optimized operating conditions aimed at reducing the fuel temperature included variation of the thermal conductivity and initial fuel-cladding gap thermal resistance. The nonlinear response along with continuum micro-cracking damage predictions are obtained and compared to the behavior during normal operating conditions. Numerical implementation of the model is based on 3-D Finite Element solutions.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Corrosion behavior of Fe–Cr–Si alloys in simulated PWR primary water environment

Fe–Cr–Si alloys are potential candidates as accident tolerant fuel (ATF) cladding, which could offer similar benefits as Fe–Cr–Al alloys. The corrosion and oxidation behaviors of three Fe–Cr–2Si alloys with various chromium content (12 wt. %, 16 wt. %, 20 wt. %) in simulated primary water chemistry of a pressurized water reactor (PWR) were investigated. Post-test characterization included weight change, scanning electron microscopy (SEM), and X-ray Photoelectron Spectroscopy (XPS). The results showed that Fe12Cr2Si had the worst corrosion/oxidation resistance among the three alloys, while Fe16Cr2Si and Fe20Cr2Si showed excellent oxidation resistance due to their thin, continuous, dense oxide layers grown in simulated PWR conditions. These results bring insight into the corrosion behavior of Fe–Cr–Si alloys with varying chromium content exposed to the primary water environment during normal operating conditions found in PWRs.

36 MATERIALS SCIENCE↗

DOE OSTI2020

D retention and depth profile behavior for single crystal tungsten with high temperature neutron irradiation

Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature. The D retention at depths up to 3 f.tm for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was increased to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W. The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Effect of heat treatment on creep behavior of 316 L stainless steel manufactured by laser powder bed fusion

The objective of this study is to understand the thermal stability of microstructure and its effect on the creep behavior of additively-manufactured 316L stainless steel (AM 316L SS). Creep specimens were fabricated from rods printed by a laser powder bed fusion process. Six different heat treatments, namely 650 °C/1 h, 700 °C/1 h, 750 °C/1 h, 800 °C/1 h, 900 °C/1 h and 1050 °C/1 h were applied to the creep specimens. The heat-treated specimens were creep-tested under the same condition, 550 °C/275 MPa to evaluate the effect of post-build heat treatment on the creep behavior of AM 316L SS. In the temperature range of 650-750 °C, dislocation density within cells, and cell size and wall thickness were affected by the heat treatment, while elemental segregation at boundaries remained unchanged and Mn-enriched Si oxide particles remained stable. In the temperature range of 750-900 °C, concurrent changes were observed in dislocation cell structures, elemental segregation and oxide particles. The 1050 °C-heat treatment removed cell structures and boundary solute segregations, leading to formation of equiaxed grains; Mn-enriched Si oxide particles in the as-built specimen were replaced by more stable Mn-enriched Cr oxides. The creep life of AM 316L SS increased after the heat treatment at 650 °C and then decreased with increasing temperature up to 900 °C where cell/subgrain structures still existed. The creep rate and creep elongation followed an opposite trend. The creep rate can be correlated with the structural parameters of cell/subgrain size and cell wall/boundary thickness and can be rationalized by different strengthening mechanisms of dislocation cells and subgrains. The heat treatment affected strongly the primary and secondary creep but had a minimal effect on the tertiary creep of AM 316L SS.

316L stainless steel↗

DOE OSTI2022

Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy

To optimize nuclear waste repository performance, the destruction of minor actinide elements, particularly Np and Am, in a neutron fast spectrum reactor is possible by incorporating these elements into nuclear fuel. Evaluating the performance of minor actinide containing fuel is of paramount importance to enabling this technology. However, such a task is challenging without an available domestic fast spectrum test reactor. A comparison of fuel performance tested in an available domestic thermal reactor at the Idaho National Laboratory, the Advanced Test Reactor, and in a fast spectrum reactor in France (Phénix) is presented here in this study. This study evaluates the capability of using a cadmium shrouded test position to mimic the power profile along the fuel radius present in fast spectrum reactors so that thermally driven phenomenon (e.g., constituent redistribution) can be evaluated in a thermal reactor and determined to be prototypical of a fast reactor. Thus, optical microscopy and scanning electron microscopy has been performed on irradiated 35U-29Pu-4Am-2Np-30Zr fuel samples (where the number preceding the element is the weight percent concentration) from the two mentioned reactors that present similar irradiation temperatures and power conditions. The results indicate that fuel performance phenomena are reproducible in the two irradiation conditions. The redistribution of Zr occurred in the same manner for the two samples. Similar partitioning of U-Pu-Zr phases was observed, and the behavior of Am was similar in the analyzed specimens. Finally, the overall microstructure evolution seems not to be affected by minor actinides addition compared to expected behavior of conventional U-19Pu-10Zr ternary metal fuels for both specimens. Slight differences in fuel cladding chemical interaction were, however, observed. This difference is likely driven by difference in cladding composition rather than irradiation conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Statistical fracture behavior of doped UO 2 using a ball-on-ring equibiaxial flexure test method

Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO 2 fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO 2 fuel fracture. UO 2 pellets were doped with 0.1 wt% TiO 2 and 0.3 wt% Cr 2 O 3 to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO 2 doping and 7 µm for Cr 2 O 3 doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO 2 doped UO 2 and undoped UO 2 , while 0.3 wt% Cr 2 O 3 doped UO 2 exhibited a 20% decrease in fracture strength. Doped UO 2 samples also showed reduced Weibull modulus compared to undoped UO 2 , suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO 2 , regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗

DOE OSTI2025

The impact of chromium coatings on Zircaloy cladding deformation behavior under reactivity-initiated accident-like mechanical loading conditions

A reactivity-initiated accident (RIA) occurs when a control rod ejection or control blade drop causes an increase in the fission rate. The injection of energy results in an increase in fuel temperature which in turn causes rapid thermal expansion of the fuel pellet. This thermal expansion may result in pellet-cladding mechanical interaction (PCMI) in which the fuel imparts a mechanical strain to the cladding. PCMI may cause the cladding to fail, and thus, the mechanical response of cladding due to PCMI must be investigated when characterizing new cladding materials. Chromium-coated Zircaloy-4 is a near-term accident-tolerant fuel cladding that exhibits improved high-temperature oxidation resistance. Modified burst testing was utilized to experimentally simulate the effects of PCMI on both uncoated and chromium-coated Zircaloy cladding samples at hot zero power conditions. Samples were coated using either cold spraying or physical vapor deposition to understand the differences in behavior that the coating application method may cause. Digital image correlation was used to analyze images of the deforming specimens to extract the in-situ strain behavior of the cladding. The uncoated specimens burst at hoop strains ranging from 8.8 % to 17.2 %. The cold-spray chromium-coated Zircaloy specimens burst at hoop strains of 7.0 % to 11.0 %. The physical vapor deposition coated tubes burst at hoop strains of 9.1 % to 11.5 %. These results indicate that the chromium coating causes a loss in the ductility of the cladding. Furthermore, the higher burst hoop strains of the physical vapor deposition-coated samples relative to the cold-spray samples indicate that the cold-spraying technique causes more of a loss in ductility than physical vapor deposition. All samples burst at higher hoop strains than those expected to occur in an RIA for fresh cladding.

Accident-tolerant fuel↗