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Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6 ± 1.6 and 5.7 ± 1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6 ± 1.1 kJ / mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models. [ABSTRACT FROM AUTHOR]

Hydrogen is a major nonmetallic impurity in the coolant of sodium-cooled fast reactors (SFRs) during normal operation. A higher hydrogen concentration than the gas–liquid equilibrium has been transiently detected in the gas space of actual SFR plants. The presence of several sodium compounds can increase hydrogen generation; however, a thorough understanding of the thermal behavior of candidate reactions is lacking. Herein, thermal analysis reveals the hydrogen release behavior of sodium hydride. Mass spectrometry indicates hydrogen generation with decreasing sample mass, indicating thermal decomposition. Detailed kinetic analysis based on master plot methods indicates that the hydrogen release reaction occurred through a mechanism involving random nucleation and growth of nuclei. Furthermore, the reaction rate was newly formulated based on a kinetic model function representing the above mechanism and the Arrhenius-type reaction rate constant comprising an activation energy of 119.0 ± 0.8 kJ mol−1 and a frequency factor of 1.8 × 107 s−1. These findings will enable the numerical simulation of the hydrogen release behavior in SFRs. [Display omitted] • Thermal analysis revealed the thermal behavior of sodium hydride as a hydrogen release source in sodium-cooled fast reactors. • Detailed kinetic analysis based on master plots showed the reaction rate and mechanism of the thermal decomposition. • A new formula was created for the reaction rate using an Arrhenius-type reaction rate constant and a kinetic model function. • Comparison of theoretical and experimental master plots confirmed a nucleation–growth-type reaction mechanism. [ABSTRACT FROM AUTHOR]

This study presents a multidisciplinary reactor-to-repository framework to compare different advanced reactors with respect to their spent nuclear fuel (SNF) disposal. The framework consists of (1) OpenMC for simulating neutronics, fuel depletion, and radioactive decays; (2) NWPY for computing the repository footprint given the thermal constraints; and (3) PFLOTRAN for simulating radionuclide transport in the geosphere to quantify the repository performance and environmental impact. We first perform the meta-analysis of past comparative analyses to identify the factors that led previously to their inconsistent conclusions. We then demonstrate the new framework by comparing five reactor types. Our analysis highlights the granularity and the specificities of each reactor and fuel type so that we should avoid making sweeping conclusions about advanced reactor SNF. Significant findings are that (1) the repository footprint is neither linearly related to SNF volume nor to decay heat, due to the repository's thermal constraint (2), fast reactors have significantly higher I-129 inventory, which is often the primary dose contributor, and (3) the repository performance primarily depends on the waste forms. The TRISO-based reactors, in particular, have significantly higher SNF volumes compared to the others but result in smaller repository footprints and lower peak dose rates. The open-source framework ensures proper multidisciplinary connections between reactor simulations and environmental assessments, as well as the transparency/traceability required for such comparative analyses. It aims to support reactor designers, repository developers, and policymakers in evaluating the impact of different reactor designs, with the ultimate goal of improving the sustainability of nuclear energy systems. [ABSTRACT FROM AUTHOR]

BODIPYs have a well‐established role in biological sciences as chemosensors and versatile biological markers due to their chemical reactivity, which allows for fine‐tuning of their photophysical characteristics. In this work, we combined the unique reactivity of arylazo sulfones with the advantages of a "sunflow" reactor to develop a fast, efficient, and versatile method for the photochemical arylation of BODIPYs and other chromophores. This approach resulted in red‐shifted emitting fluorophores due to extended electronic delocalization at the 3‐ and 5‐positions of the BODIPY core. This method represents an advantageous approach for BODIPY functionalization compared to existing strategies. [ABSTRACT FROM AUTHOR]

Because of its superior thermal-hydraulic qualities, liquid sodium has been applied to a variety of industries, including energy storage, solar energy, sodium-cooled fast reactors, and aerospace. However, fires brought on by sodium leaks at high pressure can have major thermodynamic repercussions and put employees and equipment in use at risk directly or indirectly. As a result, a realistic and accurate forecast of the combustion behavior of sodium droplet swarm can offer technical backing for the use of liquid sodium in engineering as well as a way of sodium fire prevention and control. Spray dynamics (droplet settling, droplet particle size distribution, etc.), combustion kinetics (premixed combustion, gas phase combustion, etc.), sodium aerosol diffusion, and other specialized phenomena all contribute to the complex process of sodium droplet swarm combustion. The NACOM code created by Brookhaven National Laboratory for sodium droplet swarm combustion is utilized in this paper as a framework. The code is first validated using the benchmark of the sodium droplet swarm combustion tests carried out by prestigious institutions. The validation results demonstrate that the code's drag model, droplet combustion model, and heat transfer model are to blame for the significantly overestimated thermodynamic effects of sodium droplet swarm combustion. NACOM is subsequently developed twice for the authors' previously developed vapor-liquid two-layer-structure drag model, chemical kinetic combustion model, and suitable heat transfer coefficient. It is then thoroughly assessed for the separate-effects tests and integral-effects test. The evaluation results demonstrate that the optimized drag model accelerates the settling of sodium droplets due to the consideration of the sodium-vapor drag reduction effect, reducing the thermodynamic effects of liquid sodium combustion; the optimized premixed combustion model can accurately predict the low-temperature sodium droplet swarm combustion conditions, resolving the issue of serious misvaluation of the original version of NACOM. The associated research findings can serve as valuable resources and tools for deeper comprehension of the combustion effects and mechanisms of sodium droplet swarm under various operating settings (such as leakage rate and oxygen concentration). [ABSTRACT FROM AUTHOR]

A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies. [ABSTRACT FROM AUTHOR]

The pool-type demonstrative China Fast Reactor 600 (CFR-600) adopted a series of improved safety designs, among which the hydraulic suspended passive shutdown subassembly (HS-PSS) is employed for inherent safety enhancement, especially suitable against unprotected loss of flow accident (ULOF). In this article, functional requirements for HS-PSS design in hydro-dynamic aspects are proposed, with the corresponding performance indicators discussed. To address these functional requirements, qualitative analysis on the equilibrium solution properties of the nonlinear dynamical model equation of the hydraulic moving body (HMB) in the HS-PSS are conducted, which leads to the determination of an applicable design parameter domain of the HMB for its practical design from the broad range of structural and parametric design options available for HS-PSS. Furtherly, hydraulically characterizing and modeling the constituent paths and consequent fluid network, hydraulic characteristics of the HS-PSS, as well as the coupled hydro-dynamic motion behaviors of the HMB for suspension and dropping states, were simulated and test-validated. Considering the HS-PSS hydro-dynamic behaviors, key indicators such as critical flowrate, drop time, hydraulic self-tightening performance, as well as the hydraulic characteristics curve are for fulfilling the functional requirements. Meanwhile, through sensitivity study of some structural parameters' impact on hydraulic characteristics, some most sensible structural parameters for adjusting and optimizing detailed design are observed. The work is quite significant in supporting the conceptual design of the HS-PSS as well as its engineering improvement. [ABSTRACT FROM AUTHOR]

Creep life prediction of component materials exposed to air and liquid sodium environments is critical to ensure the safe operation and structural integrity of a sodium‐cooled fast reactor. In this paper, a method for predicting the creep life of a wide range of austenitic stainless steels in air and liquid sodium was proposed based on a physics‐informed neural network. Based on the established datasets for sodium corrosion rates and creep life in air and liquid sodium, the predictive performance of physical equations, conventional machine learning models, and the proposed model were assessed. Subsequently, a data‐driven creep life assessment framework was established, providing insight into the engineering application of machine learning methods in high‐temperature structure assessment. The results show that the creep fracture of austenitic stainless steel is accelerated by liquid sodium corrosion. The proposed physics‐informed neural network exhibits enhanced suitability and accuracy for predicting the sodium corrosion rate and creep life than physical equations and conventional machine learning methods. Highlights: A data‐driven creep life prediction method for metals under corrosion was proposed.New physical constraint and feature have been proposed based on domain knowledge.The excellent predictive performance of proposed method has been validated.Data‐driven creep assessment under environment effects has been proposed. [ABSTRACT FROM AUTHOR]

Structural materials in fast reactors with harsh radiation environments due to high energy neutrons—compared to thermal reactors—potentially suffer from a higher degree of radiation damage. This radiation damage can change the thermophysical and mechanical properties of materials and, as a result, alter their performance and effective lifetime, in some cases leading to their disintegration. These phenomena can jeopardize the safety of fast reactors and thus need to be investigated. In this study, the effect of radiation damage on the vessels of molten salt fast reactors (MSFR) was evaluated based on two fundamental radiation damage parameters: displacement per atom (dpa) and primary knock-on atom (pka). Following the previous part of this article (Parts 1 and 2), an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) with stainless steel and nickel-based alloy material vessels, respectively, were considered as case studies. The SPECTER and SPECTRA-PKA codes and a PTRAC card of MCNPX, integrated with a module which has been developed in MATLAB, named PTRIM and SRIM-2013 (using binary collision approximation), were employed individually to calculate and compare dpa and PKA (this master module containing all three tools has been appended to the iMAGINE-3BIC package for future use during reactor operations). Additionally, SRIM-2013 was applied in a 3D simulation of a radiation damage map on a small sample of vessels based on the calculated PKA. Our results showed a higher degree of radiation damage in the iMAGINE vessel compared to the EVOL one, which could be expected due to the harder neutron flux spectrum of the iMAGINE core compared to EVOL. In addition, the nickel alloy vessel showed better radiation damage resistance against high energy neutrons compared to the stainless steel one, although more investigations are required on thermal neutrons and alloy corrosion mechanisms to determine the best material for use in MSFR vessels. [ABSTRACT FROM AUTHOR]

In Paper I, we presented comprehensive three-dimensional, three-component (3D, 3C) mean velocity profiles obtained through magnetic resonance velocimetry (MRV) within a 37-pin wire-wrapped rod bundle, elucidating the primary flow dynamics in sodium-cooled fast reactors. Building on this foundation, in this paper, we meticulously examine transverse velocity, helicity, and streamlines around rods, focusing particularly on the wake region of the wire within interior and edge subchannels. Our analysis uncovers a repetitive vortex evolution pattern in the wake, characterized by vortices forming, growing, weakening, and dissipating at 60° intervals of wire angular position. Furthermore, we highlight the significant variation in the azimuthal velocity component near the wire's wake, elucidating the dominant influence of wire orientation. Additionally, we identify the periodicity of flow split factor variations, demonstrating their dependence on wire orientation in interior subchannels and the prevailing impact of edge swirling in edge subchannels. Overall, our quantitatively derived experimental results provide valuable insights into flow patterns within the rod bundle, offering essential information for optimizing reactor design and analysis. [ABSTRACT FROM AUTHOR]

The sodium-cooled fast reactor (SFR), a fourth-generation nuclear reactor, utilizes liquid sodium as a coolant and offers advantages like operation at atmospheric pressure and the use of spent nuclear fuel. Understanding flow characteristics around fuel rod bundle is crucial for securing cooling performance and safety. This study investigates the flow characteristics in a 37-pin wire-wrapped fuel rod bundle of SFR from three-dimensional flow fields measured by magnetic resonance velocimetry. Aimed at enhancing the understanding of coolant flow dynamics crucial for reactor safety and efficiency, the study successfully captures and validates three-dimensional, three components of mean velocities. Key discoveries include phase differences between wire position and high velocity regions, evidence of upstream flow influence by wire structures, and secondary flows such as edge swirling induced by wire blockage. The research offers detailed insights into subchannel velocities and flow split factors, contributing to improved SFR design and safety. [ABSTRACT FROM AUTHOR]

The design of the next generation of fast breeder reactors has commenced, with the main targets being enhanced safety and improved economy. Nuclear heat generated in the fuel subassembly of fast reactors is removed by circulating sodium through the core using centrifugal pumps. The primary sodium pumps (PSPs) used are large-capacity pumps, and the design of these pumps is different from that of traditional pumps. Though many works have been reported for the performance prediction of centrifugal pumps, most of these works have been carried out in a decoupled way, and only a few works have been reported where the pump is modeled with all the associated geometric structures. Centrifugal pumps are prone to a phenomenon called suction recirculation, which occurs when pumps are operated significantly below the best efficiency point. This suction recirculation has a strong potential to damage the impeller. Correlations given in the literature for the prediction of the onset of recirculation cannot be used for complicated inlet geometries, and three-dimensional computation fluid dynamics (CFD) investigations are most suited for such applications. Many devices have been reported in the literature to reduce the intensity of (or to suppress) suction recirculation. Webs provided in the suction plenum will modify the velocity distribution at the impeller inlet and also can influence suction recirculation. In this work, the centrifugal pump used for primary sodium pumping for fast reactor applications is simulated using CFD techniques in an integrated way. The frozen rotor approach is used to simulate the impeller-diffuser hydraulics. The effect of flow hydraulics in the suction plenum, flow distribution in the standpipe–pump gap, and flow conditions in the pool on the performance characteristics of PSPs are simulated. The flow rate for the onset of suction recirculation is predicted and compared with correlations available in the literature. Simulations are carried out to study the effect of webs on suction recirculation. The effects of the number of webs and the web geometry are also studied. [ABSTRACT FROM AUTHOR]

The annular fuel has two cooling surfaces inside and outside, which can fully take away the heat of the fuel assemblies. It is urgent to conduct research on sodium-cooled fast reactor annular fuel assemblies. The Prandtl number Pr of liquid sodium is much less than 1 (Pr ˂˂ 1). The problem is that using the traditional constant turbulent Prandtl number Prt of 0.85 ~ 0.9, which will greatly affect the accuracy of the numerical prediction of the thermal-hydraulic properties of liquid sodium. In order to address this issue, the turbulent heat transfer characteristics of the bare and wire spacers 7-pin annular fuel assembly for a sodium-cooled fast reactor are investigated by a SST k-ω-kθ-εθ model. The open-source CFD software OpenFOAM is used for this calculation. In order to verify the validity of the present calculation method, some classical liquid metal heat transfer correlations are used to compare and analyze the results of the present calculation method (kθ-εθ model), the results of the Reynolds analogy hypothesis (Prt = 0.85 model), and the results of the turbulent Prandtl number correlation (Prt = kays model). It is shown that the SST k-ω-Prt = 0.85 model overestimates the Nusselt number Nu of the annular fuel assembly. The SST k-ω-kθ-εθ model agrees well with the SST k-ω-Prt = Kays model of the bare and wire spacers 7-pin annular fuel assembly. It suggests that the numerical calculation results of the SST k-ω-kθ-εθ model are reliable in both the simple and complex fuel assembly models. In brief, the SST k-ω-kθ-εθ model is more capable of predicting the turbulent heat transfer characteristics of the bare and wire spacers 7-pin annular fuel assembly in sodium-cooled fast reactors. In addition, the thermal-hydraulic characteristics of annular fuel assemblies with wire spacers for sodium-cooled fast reactors have been investigated. This research can provide a reference for the study of turbulent heat transfer characteristics of annular fuel assemblies in sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

[Display omitted] Sarcosine plays a key role in screening for early prostate cancer. However, the several already reported peroxidase mimics immobilized with sarcosine oxidase (SOX) utilized to detect uriary sarcosine still have some limitations such as complex synthesis process, using of expensive heavy metals to mimic enzyme activity and long color development time. Herein, an inexpensive peroxidase-like 2D Fe/Co-MOF nanosheet was prepared by a simple solvent modulation method. The resultant 2D Fe/Co-MOF nanosheets have strong peroxidase activity, with its Vmax value for H 2 O 2 of 15.3 × 10-8 M/s, being 1.76 times that of HRP. Then, using the 2D Fe/Co-MOF as a peroxidase model for anchoring natural SOX to construct 2D Fe/Co-MOF/SOX , which can act as a cascade reactor for detection of sarcosine. Considering the above properties, a platform for the detection of sarcosine was built based on a colorimetric method. Because of presence of the high ratio of Fe2+ caused by the electron transfer from Co2+ to Fe3+, large specific surface area and plentiful active sites, 2D Fe/Co-MOF/SOX with TMB (3,3′,5,5′-tetramethylbenzidine) colorimetric reagent could have fast color development and can be applied conveniently and fastly in an early screening tool for prostate cancer patients. The sarcosine could be quantified by peroxidase activity with a detection range of 1–400 μM and a limit of detection (LOD) of 0.324 μM. More importantly, the average sarcosine concentration of 21.367 μM and 1.871 μM was detected in patient's and normal urine (n = 5), respectively, which showed an excellent screening effect and a great potential in early prostate cancer. [ABSTRACT FROM AUTHOR]

Thermomechanical processing (TMP) of ferritic–martensitic (FM) steels, such as HT9 (Fe–12Cr–1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature. [ABSTRACT FROM AUTHOR]

Microstructured reactors offer fast chemical engineering transfer and precise microfluidic control, enabling the determination of reactions' kinetic parameters. This review examines recent advancements in measuring microreaction kinetics. It explores kinetic modeling, reaction mechanisms, and intrinsic kinetic equations pertaining to two types of microreaction: esterification and transesterification reactions involving acids, bases, or biocatalysts. The utilization of a micro packed-bed reactor successfully achieves a harmonious combination of the micro-dispersion state and the reaction kinetic characteristics. Additionally, this review presents micro-process simulation software and explores the advanced integration of microreactors with spectroscopic analyses for reaction monitoring and data acquisition. Furthermore, it elaborates on the control principles of the micro platform. The superiority of online measurement, automation, and the digitalization of the microreaction process for kinetic measurements is highlighted, showcasing the vast prospects of artificial intelligence applications. [ABSTRACT FROM AUTHOR]

The corrosion behavior and mechanism of 316 stainless steel (316SS) immersed in liquid Pb were investigated in Ar at 650 and 750 °C, respectively. The oxidation corrosion first dominated at 650 °C. The outer layer of the corroded 316SS surface was the loose iron oxide layer, and the inner layer was dense Fe-Cr spinel. However, the selective dissolution of Ni and Cr was dominant at first at 750 °C, and the O diffused to the interface increased with time, then an inner oxide layer is formed with Ni, Cr, and Fe in inner layer of the matrix. The sizes of pores formed by the dissolution gradually increased with the corrosion time increase, and the dissolved area was also partially transformed into the whole. The selective dissolution corrosion of the corroded 316SS was dominant at 750 °C with the time increase. The interaction mechanism between the liquid lead and 316SS was revealed. [ABSTRACT FROM AUTHOR]

Flow circulation and heat removal through shield and reflector assemblies can have major impacts on safety in long transients for sodium fast reactors (SFRs). These transients are typically categorized by reduced flow rates and large-scale organized flow patterns, including potential intra-assembly circulation. Such low-flow cases can provide challenges for experiments because of complications in measuring the flow rates and temperatures with high accuracy in different areas. This consequently also raises the uncertainty of many modeling approaches for these phenomena. In an effort to address some of these issues, high-fidelity large eddy simulations are performed using the highly parallel solver NekRS. A 19-pin configuration of a tight-lattice wire-wrapped hexagonal bundle (pitch-to-diameter ratio = 1.07), representing a prototypical internal configuration of a shield assembly, was investigated. The sodium flow was set at a bundle Reynolds number of 2000, with simulations being performed for modified Richardson numbers of 0.0 (i.e., no buoyancy), 0.01, and 0.04, where mixed-convection effects are anticipated. The flow and temperature fields for these cases are discussed in detail. The high-fidelity data should prove useful as reference data for expanding and improving on various reduced-resolution approaches. A basic framework for combining subchannel and computational fluid dynamics methodologies in SFRs is also presented, with preliminary results from simulations of light water reactor bundles and a discussion of changes that need to be made for potential application to SFRs. [ABSTRACT FROM AUTHOR]

The Sodium-Cooled Fast Reactor (SFR) is a promising concept chosen in the Generation IV International Forum as a possible design for pursuing the sustainable use of nuclear energy. Its core consists of multiple hydraulically isolated assemblies, with a tightly packed triangular lattice array of fuel pins enclosed in a hexagonal duct present within each assembly. Helical wire spacers are wrapped along the axis of the rods to maintain a gap between them, inducing a secondary flow, increasing the channel mixing, and enhancing convective heat transfer. In this study, a direct numerical simulation campaign is conducted for a simplified 7-pin wire wrapper geometry, with Reynolds numbers ranging from $Re$ Re = 1000 to 10 000 and a Prandtl number of $Pr$ Pr = 0.005, to investigate heat transfer in low-flow conditions. The wire wrapper case is compared to a bare bundle case with seven pins. The results are discussed, and heat transfer predictions are compared between our numerical results and classic correlations. An anisotropy invariant map is obtained for the above-mentioned cases, and turbulent kinetic energy and turbulent heat flux budgets are computed and analyzed. Our findings provide unique insights into the flow behavior within a wire-wrapped bundle. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Boron carbide (B4C)–stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B4C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B4C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B4C–liquid SS reaction based on the reduced thickness of B4C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B4C–solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B4C–liquid SS reaction, it was found that similar temperature dependency was identified between solid B4C–liquid SS and solid B4C–solid SS. Besides, the reaction rate constants of solid B4C–liquid SS were smaller than those of solid B4C–solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B4C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid–solid contact for B4C side/SS side. [ABSTRACT FROM AUTHOR]

Since F/M steel is one of the leading candidate materials for the lead-cooled fast reactor (LFR), its compatibility with the liquid LBE environment is an essential issue before application. One major way to improve LBE corrosion resistance is to control the oxygen concertation in liquid LBE for the growth of a stable, protective oxide layer on the surface of the structure material. However, the influence of the surface state on corrosion behavior is a more realistic issue when it comes to practical applications. In this study, the corrosion behavior of Si-reinforced 9Cr and 11Cr F/M steels with different surface states was investigated by a static liquid LBE corrosion test under solid-phase oxygen-controlled conditions. The result showed that at 550 °C, the coarse surface state caused dissolution behavior at the initial stage of corrosion, while the fine surface state formed the oxide layer. Moreover, at 610 °C, Si-reinforced 11Cr F/M steel shows better liquid LBE corrosion resistance due to its thinner oxide layer formation. [ABSTRACT FROM AUTHOR]

The UK government and many international experts have pointed out that nuclear energy has an important role to play in the transition towards a decarbonised energy system since it is the only freely manageable very low-carbon energy technology with 24/7 availability to complement renewables. Besides current investments in light water reactor technologies, we need innovation for improved fuel usage and reduced waste creation, like that offered by iMAGINE, for the required broad success of nuclear technologies. To allow for quick progress in innovative technologies like iMAGINE and their regulation, a timely investment into urgently needed experimental infrastructure and expertise development will be required to assure the availability of capacities and capabilities. The initial steps to start the development of such a new reactor physics experimental facility to investigate molten salt fast reactor technology are discussed, and a stepwise approach for the development of the experimental facility is described. The down selection for the choice for a diverse control and shutdown system is described through manipulating the reflector (control) and splitting the core (shutdown). The developed innovative core design of having the two core parts in two different rooms opens completely new opportunities and will allow for the manifestation of the request for separated operational and experimental crews, as nowadays requested by regulators into the built environment. The proposed physical separation of safety-relevant operational systems from the experimental room should on the one hand help to ease the access to the facility for visiting experimental specialists. On the other hand, the location of all safety-relevant systems in a now separated access-controlled area for the operational team will limit the risk of misuse through third party access. The planned experimental programme is described with the major steps as follows: core criticality experiments, followed by experiments to determine the neutron flux, neutron spectrum and power distribution as well as experiments to understand the effect of changes in reactivity and flux as a function of salt density, temperature and composition change. [ABSTRACT FROM AUTHOR]

The designed MYRRHA reactor, in its subcritical version, will be equipped with a set of detectors monitoring its condition by measuring the current value of negative reactivity, which is a crucial parameter for its safe operation. In subcritical systems, accurate and precise measurement of negative reactivity is disturbed by the so-called spatial effect, i.e., the response of detectors depends on their placement in the reactor core. This paper focuses on the Monte Carlo simulations of reactivity measurements using the area method for natU, 238U, 241Am, 239Pu, and 232Th detectors. The simulations were performed in six positions with increasing distance from the center of the core and at three axial levels. The obtained results allow for selecting optimum locations for detectors and detector nuclides in terms of the accuracy of reactivity measurement and illustrate the dependence of the reactivity on the distance. Additionally, the possibility of using 103Rh in self-powered neutron detectors was investigated. The influence of spatial effect in calculations using the area method was directly indicated in the MYRRHA reactor core for chosen isotopes and in-core positions. The results closest to true values were obtained for the second fuel assembly for 239Pu, and the third fuel assembly for natU, 238U, 232Th, and 241Am; thus, these nuclides and positions should be preferred when selecting detectors for MYRRHA. [ABSTRACT FROM AUTHOR]

In this study, we have demonstrated that data assimilation (DA) using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems (ADSs) derived from inelastic scattering cross sections of lead and bismuth. We reevaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the DA formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties and performed DA using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section–induced uncertainty of the void reactivity of the ADS from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017. [ABSTRACT FROM AUTHOR]

BACKGROUND: This study intended to detect the inherent deficiencies of prescription‐type empirical design procedures often adopted for nitrogen ‐removal activated sludge systems. Principles of process stoichiometry and kinetics were used to identify a series of hotspots that were likely to impair and disrupt system performance and efficiency. Available data, derived from an existing plant subject to a similar design approach, were utilized to numerically assess significant hotspots and to define a sustainable remedial action in compliance with applicable scientific principles. RESULTS: Critical appraisal of the implemented design exposed a large spectrum of deficiencies affecting system efficiency and performance, including nitrification efficiency, denitrification potential, anoxic volume and sludge age, internal recycling and nitrate removal. The major role of the inorganic suspended solids as a design parameter was also ignored. The remedial strategy adopted for retrofitting corrected all the observed discrepancies and it devised a sustainable approach of flipping around the existing reactors so that they could be operated as a series of carousel/fast recycle reactors, each including anoxic and aerobic volume fractions to provide full nitrification. CONCLUSION: The merit of the proposed remedial strategy was justified by offering two additional options of equally effective system operation with either reduced reactor volume or increased sewage flow. The study presented conclusive evidence to show why prescription‐type design should be avoided, providing numerical proof that this methodology is incompatible with stoichiometric and kinetic characteristics of sewage under local conditions. © 2024 Society of Chemical Industry (SCI). [ABSTRACT FROM AUTHOR]

Vibrational relaxation of H2O(v2,v13) molecules by collisions with Ar was studied at 298 K (v2 denotes the bending vibrational mode and v13 denotes the sum of the symmetric, v1, and asymmetric, v3, vibrational modes). The H2O molecules from 14 different exothermic reactions of H-atom abstraction by OH radicals were observed by infrared emission from a fast flow reactor as a function of Ar pressure and reaction time. Numerical kinetic calculations were used to obtain rate constants for stretch-to-bend energy conversion, (v2,v13) → (v2 + 2,v13 − 1), and pure bend relaxation, (v2,v13) → (v2 − 1,v13). Rate constants for states up to v13 = 4 were based on the average values from all reactions. The rate constant for the (2,0) → (1,0) bending relaxation is in agreement with the published values from laser-induced fluorescent experiments; the rate constants for higher levels increase with v2. Our average rate constant for the (0,1) → (2,0) stretch-to-bend conversion is somewhat smaller but falls within the uncertainty limit of the published value. The average rate constants for the stretch-to-bend process for (01), (02), (03), and (04) stretching states are (4.3 ± 0.8) × 10−14, (7.7 ± 1.1) × 10−14, (14.3 ± 4.2) × 10−14, and (20.6 ± 6.2) × 10−14 cm3 molecule−1 s−1, respectively. [ABSTRACT FROM AUTHOR]

Titanium-modified austenitic stainless steels (15-15TiSS) are currently used as the fuel cladding material of fast reactors, which are subject to higher radiation damage during their application. In this work, the radiation-induced swelling of 15-15TiSS under a fast reactor neutron and heavy ion irradiation conditions was studied by the rate theory (RT). The simulated swelling properties of 15-15TiSS under fast neutron conditions were calculated initially. The swelling peak, swelling rates, and swelling-related microstructural properties are consistent with the neutron irradiation results, indicating that the selected RT model and material parameters are reasonable. Then, the swelling properties of 15-15TiSS under various damage rates were predicted by changing the radiation damage rates from 1 × 10−6 to 1 × 10−3 dpa/s. It shows that swelling peaks are strongly dependent on temperature and the swelling peaks shift ∼50 °C toward the higher temperature with each order of magnitude increase of defects generation rate. The swelling rates and swelling-related defect evolution at 1 × 10−3 dpa/s (with a swelling peak temperature of 590 °C) are consistent with that under neutron irradiation with 1 × 10−6 dpa/s (with a swelling peak temperature of 460 °C). At length, the RT-predicted heavy ions irradiation results were verified by the previous positron annihilation lifetime spectroscopy results of ions irradiated 15-15TiSS. It indicates that heavy ion irradiation can be used to study the radiation effect of materials under neutron irradiation and should be a feasible technique used in the further screening of radiation-resistant materials. [ABSTRACT FROM AUTHOR]

Evaluating severe accidents in Generation IV sodium-cooled fast reactors (SFRs) is challenging due to the eutectic reaction between boron carbide (B 4 C) and stainless steel (SS), which results in boron migration and enhanced neutron absorption. A novel quantitative visualization method using radiative heating to observe eutectic behavior and resulting melt structure during boron migration. Experiments, replicating real-scale B 4 C-SS control rods, were conducted in an integrated test facility under core disruptive accidents (CDAs) conditions, introducing a new approach. This study explores the long duration melting (candling phenomenon) of boron within stainless steel, shedding light on relocation behavior not previously studied. We identified two distinct failure mechanisms: the separation of SS from the B 4 C pellet, resulting in the formation of a later melting drop, and the fracture of the B 4 C pellet into multiple pieces, possibly due to thermal stress. The visualization technique and thermal interfacial resistance analysis precisely captured the eutectic temperature, aligning well with prior research. Our study yield insights into eutectic melt relocation behavior by systematically characterizing relocated solidified eutectic melt with a cooling rate of 50 °C/s in stagnant argon environment via scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDS), X-ray diffraction (XRD), and microhardness testing. Our study shows that due to high chromium content, (Cr,Fe) 2 B phases are observed in XRD analysis instead of just Fe 2 B phase. Other phases present in our solidified eutectic melt include γ-Fe (fcc), (Cr,Fe) 23 (B,C) 6 (cubic), and (Cr,Fe) 5 B 3. The atomic composition within these zones was ascertained through SEM-EDS analysis. Our study shows varying hardness values in distinct phases, reflecting differing boron concentrations within micrograph zones. Chromium promotes (Cr,Fe) 2 B and (Cr,Fe) 5 B 3 boride formation, transforming the structure from tetragonal to orthorhombic with increased chromium content, thereby increasing hardness. This validates B 4 C-SS eutectic mixture relocation and diverse boride phase formation under extreme reactor conditions. [ABSTRACT FROM AUTHOR]

The computational demand of neutron Monte Carlo transport simulations can increase rapidly with the spatial and energy resolution of tallied physical quantities. Convolutional neural networks have been used to increase the resolution of Monte Carlo simulations of light water reactor assemblies while preserving accuracy with negligible additional computational cost. Here, we show that a convolutional neural network can also be used to upsample tally results from Monte Carlo simulations of sodium-cooled fast reactor assemblies, thereby extending the applicability beyond thermal systems. The convolutional neural network model is trained using neutron flux tallies from 300 procedurally generated nuclear reactor assemblies simulated using OpenMC. Validation and test datasets included 16 simulations of procedurally generated assemblies, and a realistic simulation of a European sodium-cooled fast reactor assembly was included in the test dataset. We show the residuals between the high-resolution flux tallies predicted by the neural network and high-resolution Monte Carlo tallies on relative and absolute bases. The network can upsample tallies from simulations of fast reactor assemblies with diverse and heterogeneous materials and geometries by a factor of two in each spatial and energy dimension. The network's predictions are within the statistical uncertainty of the Monte Carlo tallies in almost all cases. This includes test assemblies for which burnup values and geometric parameters were well outside the ranges of those in assemblies used to train the network. [ABSTRACT FROM AUTHOR]

Herein, a science‐based uranium and plutonium mixed oxide (MOX) property model is derived for determining the properties of MOX fuel and analyzing their performance as functions of Pu content, minor‐actinide content, oxygen‐to‐metal ratio, and temperature. The property model is constructed by evaluating the effect of phonons and electronic defects on the heat capacity and thermal conductivity of MOX fuels. The effect of phonons was evaluated based on experimental datasets related to lattice parameters, thermal expansion, and sound speeds. Moreover, the effect of electronic defects was determined by analyzing oxygen‐potential data based on defect chemistry. Furthermore, the model evaluated the effect of the Bredig transition on the thermal properties of MOX fuel by analyzing the irradiation test results. The derived property model is applied to the performance code to analyze fast reactor fuel pins. [ABSTRACT FROM AUTHOR]

Lead-cooled fast reactors exhibit strong inherent safety performance and good economic features, while material degradation due to corrosion and irradiation is still challenging. Oxide dispersion-strengthened steels are one of the promising candidates for fuel cladding materials. The effects of both irradiation and corrosion on ODS steel need to be further studied. In this work, MX-ODS steel was irradiated by Fe ions at 500 °C up to 46 dpa. Later, the as-received specimen and the irradiated specimen were used to conduct corrosion tests in oxygen-saturated Pb at 550 °C for 1 h. In the as-received specimen, discontinuous oxides penetrated by Pb and Pb in contact with steel matrix were observed, demonstrating unsatisfactory corrosion resistance of the material. However, in the irradiated specimen after corrosion experiment, a protective oxide layer formed and prevented Pb attack. The oxidation behavior differences between the two specimens can be attributed to the defects produced by irradiation and the structural discrepancy in oxides caused by the formation process. A possible mechanism of irradiation on the corrosion is discussed. In the as-received specimen, Fe atoms loss led to voids in the oxides, and lead penetrated the oxides through these voids. In the irradiated specimen, defects left by previous irradiation helped to form a more uniform oxide layer. The adhesive outer magnetite oxide and the Fe ions generated from where grain boundary oxidation developed retarded the presence of voids and made the oxide layer protective. [ABSTRACT FROM AUTHOR]

This paper introduces a novel concept of a TRISO-fueled salt-cooled reactor (TFSCR). The core contains circulating pipes filled with molten salt carrying TRISO particles. The reactor achieves online refueling by slowly circulating the molten salt through the pipeline. The reactor utilizes the same molten salt as the coolant and graphite as the moderator. The reactor design has the characteristics of safety, economy, and nonproliferation. TRISO particles exhibit greater resistance to neutron irradiation, corrosion, oxidation, and high temperatures compared to conventional fuels. A molten salt-cooled reactor can also operate at higher temperatures, consequently enhancing power generation efficiency. Furthermore, lower operating pressures can mitigate the risk of significant damages and loss of coolant caused by accidents, thereby enhancing reactor safety. This paper presents the basic nuclear design of TFSCR under the assumptions concerning the average fuel power density, the volumetric core power density, and the core temperature. At the same time, the feasibility of online refueling and long-life operation was evaluated by fuel burnup calculation. [ABSTRACT FROM AUTHOR]

Shell and tube-type crossflow heat exchangers with circular fins are widely used in various industries, including nuclear power plants. This study focuses on the sodium-to-air heat exchanger, which is one of the key safety features in sodium-cooled fast reactors. Tube alignment inside the heat exchanger is important because it directly determines the performance. Although heat exchangers are generally designed to have staggered tube alignment, inevitable inline-aligned zones exist. Therefore, this paper proposes a hybrid tube alignment modeling approach that considers both staggered and inline tube heat transfer phenomena. The results of the hybrid model calculations are verified and validated with liquid sodium experiments. The hybrid approach to the tube arrangement reduced the temperature deviation between experimental and calculated values to a difference of 4.55% and 7.38% for both sodium and air sides, respectively, while for the heat transfer, the difference was 15.12% and 9.06% for sodium and air sides, respectively. The results of this study can also be used as a basis for the safety evaluation of nuclear reactors and licensing processes for sodium-cooled fast reactors. In addition, this study is limited not only to sodium heat exchangers but also to any serpentine tube arrangement or crossflow heat exchanger under high-temperature systems, such as concentrated solar power and thermal energy storage industries. [ABSTRACT FROM AUTHOR]

Double-layer heat tubes have been designed to effectively reduce the occurrence of heat pipe rupture accidents. However, inter-tube thermal contact resistance can decrease heat transfer efficiency, thus hampering the heat dissipation in the primary loop system of lead-bismuth reactors. Therefore, optimizing the design of double-layer heat tubes is necessary. This work focuses on the double-layer heat exchanger of a lead-bismuth reactor and utilizes gallium-based graphene nanofluids as a thermal interface material to fill the gap between the heat tubes. Furthermore, the impact of the length, wall thickness, outer diameter, and spacing of heat tubes on the heat transfer performance of the double-layer heat exchanger with and without the nanofluids has been analyzed. The study aims to optimize the JF factor and cost-effectiveness ratio (CER). Genetic algorithms are employed to optimize and evaluate the heat transfer performance of the main heat exchanger based on the four aforementioned parameters. Consequently, a new design scheme is obtained for the double-layer heat exchanger, which increases the optimized overall heat transfer coefficient of the main heat exchanger by 5.79%, pressure drop in the primary loop by 2.32%, JF factor by 5%, and CER by 24.62%. These results demonstrate that the gallium-based graphene nanofluids can effectively enhance the heat transfer performance of the double-layer heat exchanger while reducing the likelihood of steam generator tube rupture accidents. [ABSTRACT FROM AUTHOR]

The finned-tube sodium-to-air heat exchanger plays a crucial role in ensuring the safety of prototype Gen IV sodium-cooled fast reactors. It is designed to actively operate with power, but it is also important to consider its passive performance to ensure availability during power-out conditions. This study focused on addressing this limitation through a natural convection experiment using a dedicated facility, and the obtained results were compared with the safety analysis code MARS-LMR. The overall temperature difference was observed to be within an acceptable range of less than 5 %. While the code calculations aligned well with the experimental results at higher temperatures, they underestimated the experimental results at lower temperatures. These findings can serve as a basis for other codes employing different heat transfer correlations and for future reactor licensing purposes. [ABSTRACT FROM AUTHOR]

Aim: To assess and compare the metabolic and vascular effectiveness of sodium‐glucose cotransporter 2 inhibitors (SGLT‐2i) and dipeptidyl peptidase‐4 inhibitors (DPP‐4i) in the clinical practice of patients with type 2 diabetes in Italy. Materials and Methods: GIOIA is a 2‐year prospective, multicentre, quasi‐experimental study that enrolled patients with type 2 diabetes initiating SGLT‐2i or DPP‐4i for inadequate glycaemic control [glycated haemoglobin (HbA1c) >7%] between March 2018 and March 2021. The primary endpoints were changes in markers of organ damage [carotid intima‐media thickness (CIMT), albuminuria, myocardial function] and HbA1c from baseline to year 2. Results: In total, 1150 patients were enrolled in the study (SGLT‐2i n = 580, DPP‐4i n = 570). Patients initiated on SGLT‐2i were younger (about 6 years) and heavier (about 11 kg), had higher HbA1c level (1% more), more albuminuria and cardiovascular events (16% more) than patients initiated on DPP‐4i. CIMT and echocardiographic parameters were not significantly different. Propensity score matching yielded two groups, each consisting of 155 patients with diabetes with similar baseline characteristics. Despite a significant similar reduction in HbA1c levels in both groups (−0.8%), more patients on SGLT‐2i had regression of CIMT and albuminuria (22% and 10%, respectively, p <.001 vs. DPP‐4i); more patients on DPP‐4i had progression of CIMT and albuminuria (23% and 28%, respectively, p <.001 vs. SGLT‐2i). Left ventricular ejection fraction improved slightly (3%, p =.043) on SGLT‐2i only. Conclusions: In a real‐world setting, both SGLT‐2i and DPP‐4i improve glycaemic control persisting after 2 years of treatment, with a robust effect on both CIMT and albuminuria regression for SGLT‐2i as compared with DPP‐4i in the propensity score matching. [ABSTRACT FROM AUTHOR]