Accident Tolerance Research Articles

Enhanced accident tolerance properties of AlCrCuFeMoNbx high entropy coating for nuclear fuel cladding

High entropy alloy AlCrCuFeMoNbx (x=0.5, 0.8, 1.4, 2.0) coatings were deposited on Zr-1Nb alloy substrates by magnetron co-sputtering to improve the accident tolerance capability of the nuclear fuel cladding. The structural and accidental tolerant properties of high-entropy alloy coatings have been investigated comprehensively. The results show that coatings with varying Nb contents exhibit amorphous structures. The hardness and elastic modulus of the coatings increase with the Nb content, reaching their maximum values at x = 2.0, which are 11.38GPa and 186.05GPa, respectively. In both the simulated normal operating and accident environment of the nuclear fuel cladding, the oxidation resistance of the coating has been investigated. The best oxidation resistance was achieved at x=0.8 in pure water at 360℃under 18.6MPa for 3 days. The dense Cr2O3 and CrNbO4 oxide films generated on the surface hinder the diffusion of oxygen ions, and account for the lower weight gain 5.10mg/dm2, decreased by 56.4% compared to that of the uncoated zirconium alloy. The coating with x=0.5 showed a drastic reduction in oxidized weight gain by 84.02% compared to the uncoated zirconium alloy in water steam at 1200℃, demonstrating excellent accident tolerance, which is due to the dense α-Al2O3 and CrNbO4 protective layer generated on the surface of the coating. These findings suggest that high entropy alloy coatings have promising accidental tolerance capabilities for nuclear fuel cladding.

Hoop tensile properties and crack propagation investigation of 2D braided SiCf/SiC composite tubes: Experiments and simulations

SiCf/SiC composites are promising candidates for advanced pressurized water reactors (PWRs) fuel cladding materials due to their enhanced accident tolerance. Their mechanical properties are strongly influenced by the braided structure of continuous SiC fibers. This study fabricated SiCf/SiC composite tubes with braid angles ranging from 30° to 50°, evaluating their hoop tensile properties through expansion-due-to-compression (EDC) experiments and analyzing the damage process using finite element method simulation. Results indicate that variations in braid angles significantly affect structural density, thereby impacting mechanical strength under hoop tensile stress. Increased braid angles result in smaller pore units and higher pore density, leading to local stress concentrations and varied deflections at overlapping regions. The dynamic propagation behavior of cracks was investigated through acoustic and structural nondestructive testing methods. Finite element analysis of different braid configurations highlights the pivotal role of pore units in the initiation and propagation of hoop tensile cracks. This study enhances the understanding of toughening structures in 2D braided composites and provides a theoretical basis for future accident-tolerant fuel cladding design.

Corrosion of Chromium Coating Fabricated on Zircaloy-4 Substrate.

In the nuclear industry, coated cladding is a topical problem and it is chosen as the near-term and most promising ATF (Accident-Tolerant Fuel) cladding concept. The main objective of this concept is to enhance the accident tolerance of nuclear power plants and accordingly, the performance of cladding is expected to be improved. This work assesses the corrosion performance of a Zircalloy-4 alloy coated with a thin chromium coating by MS (magnetron sputtering), tested under a CANDU (CANada Deuterium Uranium) reactor primary circuit simulated condition (LiOH solution, 10 MPa, 310 °C, pH = 10.5). The anticorrosive performance is evaluated by a gravimetric analysis, a metallographic analysis, X-ray diffraction, electronic microscopy, and electrochemical methods. A four times less gain mass was noticed compared to uncoated Zircaloy-4, indicating a smaller corrosion rate. The SEM micrographs illustrate that the coatings are still adherent, and they are keeping the initial morphological characteristics during the autoclaving process. A SEM cross-section analysis shows values of the thickness of the coatings between 0.8 and 1.46 µm. By XRD, the presence of Cr2O3 oxide is identified. Electrochemical testing confirms good stability and good corrosion performance of Cr coating over time under autoclave conditions.

Neutron-physical characteristics of UO2 and UN/U3Si2 fuels with Zr, SiC and APMT accident tolerant claddings

H2 production due to zirconium hydration was the primary source of explosion in the Fukushima Daiichi nuclear accident. To improve accident tolerance of the new reactor designs, advanced fuel and cladding materials are being extensively investigated. Neutronic evaluation at both the pin-cell and assembly levels are performed for the conventional UO2 and the candidate UN/U3Si2 fuels with traditional cladding (Zr) as well as the accident tolerant claddings (SiC and advanced powder metallurgic ferritic, APMT). The neutronic performance, plutonium inventory, cycle length, radial and pin power distribution, spectrum distribution, spatial self-shielding analysis, plutonium radial distribution, and reactivity coefficients are evaluated and analyzed. Combination of UN/U3Si2 and APMT produces 68% and 66 % more 239Pu and 135Xe than the UO2 and APMT systems. In overall, the reactivity change versus burnup and other main neutronic parameters of SiC cladding are quite similar to the current Zr cladding. Using SiC resulted in an average increase of 0.85 % in reactivity in comparison to Zr conventional cladding (without any change in the enrichment of the fuel). Compared with the traditionally used UO2 fuel assembly, the UN/U3Si2 fuel assembly exhibit a higher (0.2 % at BOC and 0.4 % at EOC) pin power peak value in an assembly, which should be a concern from safety aspects. The negativity of FTC and MTC for the proposed UN/U3Si2 is higher than that of UO2 fuel. These higher and enhanced FTC and MTC values of UN fuels mean an improved inherent safety feature of UN-like high U density fuels.

The effect of nanometre-scale kinetic competition on the phase selection in Zr/Si superstructure

Nano structured coating system demonstrates outstanding performance stemming from unique structure evolution. Zr/Si coating was proposed as an ideal candidate for the protection of Zr alloy cladding. However, the phase selection remains a challenge to coordinate the complex requirement under the service and accident conditions of reactor core. In this work, a phase selection mechanism for amorphous Zr-Si-O (abbr. ZSO) was elucidated by the competitive consumption of Zr layer by silicidation and oxidation, which was also scale-dependent. The energy-loss near-edge fine structure and nanoindentation were also employed to characterize the unique scale effect in the two processes. Both the oxides kinetics and elastic modulus gain indicated an immediate passivation process under the subcritical condition after the in situ phase selection of the amorphous ZSO. Ab initio molecular dynamics (AIMD) was performed to disclose the origination of depressed O migration in ZSO. The multilayer structure also shows accident tolerance potential. The interfacial phase selection provided a strategy to regulate mesoscale structure originating from nanomter interface and meets the diversified requirement for “structure-performance relationship” under the service and accident conditions.

Core design: Review & perspectives in Spain

In Core Fuel Management (ICFM) is an essential part of the operation of all commercial nuclear reactors. At the beginning of each fuel cycle (BOC), an optimal loading pattern is required to produce safe and economical operating conditions, such as uniform power distribution and a high degree of fuel burnup. These operating conditions are sought to achieve maximum useful life at minimum cost and the highest levels of safety, which leads to maximization of benefits. This article describes the work done by Spanish nuclear related organizations, which are currently involved in core design and core analyses from industrial, research and safety perspectives. In many instances, achieving enhanced accident tolerance would provide significant opportunities for outstanding research and improvements in operational cost benefits for the industry.

Multiphysics fuel performance modeling of dispersed TRISO-coated particle fuel plate under long-time normal condition and accident conditions

TRISO-coated nuclear fuel has been extensively studied these decades due to the reliable accident tolerance performance. Based on the ABAQUS implicit finite element solver, the thermal–mechanical coupled simulations of the TRISO-particles embedded plate fuel were carried out. Multi-physical effects were incorporated into the simulation, including the plastic deformation, irradiation swelling, creep and gap heat transfer. Results show that the dispersed TRISO-coated particle fuel plate has robust thermal–mechanical performance under long-time normal condition. The maximum fuel temperature is 668 K and the cumulative failure probability for the SiC layer is 1.85 × 10−4. Under reactivity insertion accident and loss of coolant accident conditions, this fuel has reliable anti-accident capacities due to TRISO layer’s protection. Additionally, the porosity effect on thermomechanical performance of this fuel was analyzed.

Impingement wastage experiment with SUS 316 in a printed circuit steam generator

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

Displacement cascade under high-energy irradiation and tensile property in Fe-Cr-Al alloys: A molecular dynamics simulation

Fe-Cr-Al alloy is considered as one of the most prospective accident-tolerant fuel cladding materials due to its excellently comprehensive properties. Irradiation performance evaluation of Fe-Cr-Al alloy is the key for successful development of fuel-clad system with enhanced accident tolerance. However, the microstructural response is largely dependent on irradiated conditions. In the present work, molecular dynamic method is employed to simulate the cascade collision processes under high-energy irradiation and tensile deformation behaviors of Fe-Cr-Al alloys. The threshold displacement energies (TDEs) in Fe-Cr-Al alloys are calculated at different temperatures and along various directions, and the TDEs along different directions exhibit significant anisotropy. The effect of alloy compositions on the displacement cascade of Fe-Cr-Al alloys is systematically studied, as well as the direction of the primary knock-on atom (PKA), temperature, and PKA energy. At 300 K and 700 K, no significant dependence between the number of surviving Frenkel pairs and PKA energies is observed in Fe-Cr-Al alloys under high-energy irradiation. High temperature and high-energy PKA could potentially lead to form large-sized interstitial clusters and a few small-sized vacancy clusters in Fe-Cr-Al alloys. Dislocations or dislocation loops can easily form at high temperatures and in high-energy irradiation. The effect of alloy compositions, temperatures and PKA energies on the tensile properties of Fe-Cr-Al alloys before and after irradiation, is systematically evaluated. By analyzing the corresponding yield strength, ultimate tensile strength and elastic modulus, Fe-Cr-Al alloys exhibit the characteristic of irradiation-induced hardening and embrittlement with the increase of PKA energy. As the temperature increases, the rigidity of alloys decreases while the irradiation embrittlement significantly increases.

In-situ observation of damage evolution in Mo-SiCf/SiC heterogeneous composite

Innovative metal-SiCf/SiC heterogeneous structure composites are promising accident tolerance fuel cladding materials for the Gen IV advanced reactors. The mechanical damage of these composites is a critical issue for their safe application. In this work, the damage process and mechanism of the Mo-SiCf/SiC cladding are revealed by combining In-situ C-ring test and Finite Element Analysis. The results show that the mirco-cracks originate from the pores in the outermost side of the SiCf/SiC layer, and propagate along the fiber bundles boundaries. At the same time, the “bamboo joint”-like damage occurs in the SiCf/SiC layer during C-ring test due to the periodic pore distribution. The fact that the formation of the main crack in SiCf/SiC layer prior to the fracture of Mo layer indicate the Mo-SiCf/SiC cladding can maintain the hermeticity and structure integrity until the sample fails completely. Moreover, the C-ring strength of the Mo-SiCf/SiC cladding is 360 MPa, which is much higher than 201 MPa of SiCf/SiC cladding. Overall, the Mo-SiCf/SiC cladding structure exhibits better crack growth resistance and higher strength compared to SiCf/SiC cladding. Meanwhile this work will provide technical support and theoretical basis for the structural design of cladding materials for advanced nuclear energy systems.

Investigations on mechanical, thermodynamic and surface properties of U–Si alloys

The investigation of U–Si alloys as advanced technology fuels or accident tolerance fuels (ATFs) has gained momentum due to their higher uranium density and exceptional thermal conductivity. Density functional theory (DFT) calculations were used to examine the mechanical, thermodynamic and surface properties of U–Si alloys. Voigt-Reuss-Hill model was used to evaluate the mechanical properties such as Young's modulus, shear modulus, bulk modulus and so on. The results suggest that U3Si2 possesses an inherent brittleness, as opposed to the remaining U–Si alloys which demonstrate reliable ductility. In comparison to the hardness of other U–Si alloys (40–80 GPa), U3Si2 exhibits a significantly lower hardness (∼10 GPa). The results of anisotropy show that U3Si2 exhibits a lower anisotropy compared to USi2 (I41/amd) and U3Si (Pm-3m). Also, based on these mechanical properties, the sound velocity and Debye temperature are determined. Additionally, the study calculates the internal energy, Helmholtz free energy, entropy and heat capacity as a function of temperature, which is beneficial for fuel design and other applications. The surface energies and work function are investigated comprehensively for 161 surface slab models and used to obtain other surface properties, including equilibrium morphology, dominant surface orientations, area-weighted surface energy and area-weighted work function.

The surface and grain boundary properties of uranium boride: A DFT calculation

Uranium borides are currently being investigated as an accident tolerance fuel or advanced technology fuels (ATFs) to replace conventional UO2 fuel due to their high uranium density and high thermal conductivity. The surface properties of diverse facets of a crystal are pivotal for understanding different corrosion behaviors, fission gas bubble behaviors, and equilibrium morphologies. In this study, we employ density functional theory (DFT) calculations to investigate the surface properties of uranium borides. The surface orientations with maximum miller index up to 4, 1, and 2 for UB2, UB4 and UB12 were analyzed, respectively. Based on the calculated surface energies, the other surface properties of uranium borides were obtained through Wulff construction, including equilibrium morphology, dominant surface orientations, area-weighted surface energy and work function. Additionally, grain boundaries (GBs) have a significant impact on material properties, such as mechanical properties, corrosion behavior, and resistance to irradiation. Therefore, we use DFT calculations to determine the GB energies and work of separation for three low-Σ (7) symmetrical tilt GBs (STGB) and two twist GBs (TGB).

Best estimate plus uncertainty analysis of metal-water reaction transient experiment

Uncertainty analysis is applied in the licensing process for nuclear installations to complement best estimate analysis and to verify that the upper bound value is less than the threshold corresponding to the safety parameter of interest. Metal-water reaction is a critical safety phenomenon of water-cooled nuclear reactors at accident conditions, e.g. Loss-Of-Coolant Accidents (LOCA). AISI 348 cladding is able to increase the accident tolerance comparing to Zr-based alloys and differently from other accident tolerant fuel cladding options, there is operational experience of nuclear power plants with stainless steel. In this study, a transient oxidation experiment of AISI 348 by steam was conducted and the major sources of uncertainty were addressed. An evaluation model was developed to calculate the evolution of mass gain during the experiment. Meanwhile, uncertainty propagation of experimental data was performed. The results show that the mass gain predicted by the transient metal-water reaction model lays within the experimental data uncertainty band. Furthermore, the selection of the oxidation kinetics model seems to be important whether the analysis wills to provide conservative results.

The effects of vacancies and fission products on the structures and properties of β-SiC: a first-principles study

Understanding the interaction between fission products (FPs) from nuclear fuel nuclear and SiC is important for SiCf/SiC composites as an accident tolerance fuel cladding (ATFC). Thus, the effects of vacancies and FPs (I/Pd/Xe) on the structure, mechanical and thermal properties of β-SiC were investigated by DFT calculations in this work. The formation of defects (vacancies, FPs) could result in the distortion of SiC lattice and the loss of symmetry. The calculated binding energies indicated that most vacancies and FPs atoms could attract each other to form complex defects (vacancies clusters or FPs on vacancy sites). Mechanical calculations indicated that the formation of defects inside SiC led to the degradation of its mechanical properties. The more vacancies, the lower elastic moduli and tensile strength. While the effects of FPs on the elastic moduli were complex. Further tensile calculations indicated that Xe led to tensile strength degrade most significantly because of its weak interaction with surrounding atoms. For thermal properties, considering the vacancies only, isobaric heat capacity and thermal conductivity of SiC reduced while thermal expansion increased. In addition, the I/Pd/Xe could lead to an increase in thermal expansion and heat capacity but a decrease in thermal conductivity. In most case, Xe had the greatest effect on thermal properties.

High‐temperature oxidation behavior of the SiC layer of TRISO particles in low‐pressure oxygen

Surrogate tristructural-isotropic (TRISO)-coated fuel particles were oxidized in 0.2 kPa O2 at 1200–1600°C to examine the behavior of the SiC layer and understand the mechanisms. The thickness and microstructure of the resultant SiO2 layers were analyzed using scanning electron microscopy, focused ion beam, and transmission electron microscopy. The majority of the surface comprised smooth, amorphous SiO2 with a constant thickness indicative of passive oxidation. The apparent activation energy for oxide growth was 188 ± 8 kJ/mol and consistent across all temperatures in 0.2 kPa O2. The relationship between activation energy and oxidation mechanism is discussed. Raised nodules of porous, crystalline SiO2 were dispersed across the surface, suggesting that active oxidation and redeposition occurred in those locations. These nodules were correlated with clusters of nanocrystalline SiC grains, which may facilitate active oxidation. These findings suggest that microstructural inhomogeneities such as irregular grain size influence the oxidation response of the SiC layer of TRISO particles and may influence their accident tolerance.

Improved irradiation resistance of Cr-Fe alloy for Cr-coated Zircaloy application in accident tolerant fuel

Cr coating onto Zircaloy for the application of accident tolerance fuel claddings became a popular topic in the nuclear material field since the Fukushima-Daiichi accident in 2011. However, in the Cr coating layer the solute element effect on the potential performance improvement is still unknown, especially under irradiation. The irradiation effect in Cr-7Fe was studied after irradiation by 2.8 MeV Fe2+ to 20 dpa at 550 °C. The nano-indentation results show that doping of Fe not only can strengthen the material but also can significantly reduce the irradiation-induced hardness compared with pure Cr. The microstructure analyses show that solute element Fe can retain the diffusion of defects and reduce void swelling. In addition, the APT results indicate that no detrimental Fe-rich cluster formed in irradiated Cr-7Fe. From the aspects of irradiation-induced hardening, void-swelling, and clustering, a great improvement in irradiation resistance was achieved in Cr-7Fe alloy compared with pure Cr.

Structural Evolution and Transitions of Mechanisms in Creep Deformation of Nanocrystalline FeCrAl Alloys.

FeCrAl alloys have been suggested as one of the most promising fuel cladding materials for the development of accident tolerance fuel. Creep is one of the important mechanical properties of the FeCrAl alloys used as fuel claddings under high temperature conditions. This work aims to elucidate the deformation feature and underlying mechanism during the creep process of nanocrystalline FeCrAl alloys using atomistic simulations. The creep curves at different conditions are simulated for FeCrAl alloys with grain sizes (GS) of 5.6-40 nm, and the dependence of creep on temperature, stress and GS are analyzed. The transitions of the mechanisms are analyzed by stress and GS exponents firstly, and further checked not only from microstructural evidence, but also from a vital comparison of activation energies for creep and diffusion. Under low stress conditions, grain boundary (GB) diffusion contributes more to the overall creep deformation than lattice diffusion does for the alloy with small GSs. However, for the alloy with larger GSs, lattice diffusion controls creep. Additionally, a high temperature helps the transition of diffusional creep from the GB to the dominant lattice. Under medium- and high-stress conditions, GB slip and dislocation motion begin to control the creep mechanism. The amount of GB slip increases with the temperature, or decreases with GS. GS and temperature also have an impact on the dislocation behavior. The higher the temperature or the smaller the GS is, the smaller the stress at which the dislocation motion begins to affect creep.

The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. In this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO2 formation via Cr2O3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al2O3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H2 generation by orders of magnitude relative to bare Zry-4.

Assessment of neutronics performance of U3Si, U3Si2, and U–9Mo fuels for the support of MNSR conversion using SCALE 6.2.3 computational tools

In spite of the acceptability and production of Uranium Oxides (UO2) with Low Uranium Enrichment (LEU-UO2) fuel in commercial scale, the use of uranium–silicide (U–Si) in place of the UO2 is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. In this study, SCALE 6.2.3 computational tools have been used to model the Nigerian Research Reactor-1 (NIRR-1) which is a typical Miniature Neutron Source Reactor (MNSR) and calculated the core neutronic parameters such as Clean Core Cold Excess Reactivity (CCER), Control Rod Worth (CRW), Shutdown Margin (SDM), Safety Reactivity Factor (SRF) and Neutron flux distribution for the Highly Enriched Uranium (HEU) and candidates LEU cores. The results showed a close agreement between the calculated and measured data. The bias between calculated and measured data was attributed to the computational limitation of this study that was caused by a limited number of particles histories used in criticality calculations. The neutronic performance of the U3Si2–Al, U3Si–Al and U9Mo–Al fuels indicated the suitability of these candidates LEU fuels for conversion of MNSRs from HEU to LEU. The observed reduction in the thermal flux of the candidate LEU fuels was attributed to higher inventory of U-238 in LEU fuels when compared with the HEU and further underscore the need to raise their thermal power by approximately 10% for efficient utilization of the MNSR in Neutron Activation Analysis.

Multi-modal tomographic imaging system for poolside characterization of nuclear test fuels: Design considerations and studies

Testing and qualification of advanced nuclear fuels involve an iterative process of prototyping, in-pile irradiation testing, and in/ex situ examination. Fuel restructuring and fission product migration during burnup are among the most important aspects of fuel evolution and affect several important performance characteristics (e.g., heat removal, accident tolerance, and fission product retention). Poolside nondestructive characterization techniques provide fuel developers with tools to understand fuel evolution at different time points of burnup. A design for a compact, submersible, and multi-modal (transmission and emission) gamma-ray tomography system for imaging irradiated nuclear fuel is presented herein. Detector selection, collimator geometry and fabrication, mechanical design, imaging protocol, and acquisition protocol are discussed. Modeling calculations showed that sub-millimeter resolution could be achieved in a matter of hours in both transmission and emission tomography images. Several design compromises and fabrication challenges are discussed to further the development of future submersible gamma-ray tomography instruments.