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Multi-principal component transition metal (TM) diborides represent a class of high-entropy ceramics (HECs) that have received considerable interest in recent years owing to their promising properties for extreme environment applications that include thermal/ environmental barriers, hypersonic vehicles, turbine engines, and next-generation nuclear reactors. While the addition of chemical disorder through the random distribution of TM elements on the cation sublattice has offered opportunities to tailor elastic stiffness and hardness, the effects of irradiation-induced structural damage on the physical properties of these complex materials have remained largely unexplored. To this end, changes in the hardness and elastic moduli of a high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 and three of its quaternary subsets following irradiation with 10 MeV gold (Au) ions to fluences of up to 6 × 1015 Au cm−2 are investigated at the micrometer and sub-micrometer length-scales via the dispersion of laser-generated surface acoustic waves (SAW) and nanoindentation, respectively. The nanoindentation measurements show that the TM diborides exhibit an initial increase in hardness following irradiation with energetic Au ions, with a subsequent decrease in hardness following further irradiation. One quaternary composition, (Hf1/3Ta1/3Ti1/3)B2, exhibits a notable exception to the trend and continues to exhibit an increase in hardness with ion irradiation fluence. Although differences in the absolute values of the effective elastic moduli obtained from the measured SAW dispersion and nanoindentation are observed (and attributed to microstructural variations at the measurement length-scale), both techniques yield similar trends in the form of an initial reduction and subsequent saturation in the elastic modulus with increasing ion irradiation fluence. The quaternary TM diboride (Hf1/3Ta1/3Ti1/3)B2 again exhibits a departure from this trend. The high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 exhibits the greatest recovery in hardness and modulus when irradiated to high ion fluences following initial changes at low fluence, indicating superior resistance to radiation-induced damage over its quaternary counterparts. Opportunities for designing HECs with superior hardness and modulus for enhanced radiation resistance (compared to their single constituent counterparts) by tailoring chemical disorder and bond character in the lattice are discussed. [ABSTRACT FROM AUTHOR]

We provide a concurrent measurement of the hydrogen and oxygen nuclear kinetic energies in the water molecule across melting at 270 K in the solid phase and 276 K in the liquid phase. Experimental values are obtained by analyzing the neutron Compton profiles of each atomic species in a deep inelastic neutron scattering experiment. The concurrent measurement of the atom kinetic energy of both hydrogen and oxygen allows the estimate of the total kinetic energy per molecule due to the motion of nuclei, specifically 35.3 ± 0.8 and 34.8 ± 0.8 kJ/mol for the solid and liquid phases, respectively. Such a small difference supports results from ab initio simulations and phenomenological models from the literature on the mechanism of competing quantum effects across the phase change. Despite the experimental uncertainties, the results are consistent with the trend from state-of-the-art computer simulations, whereby the atom and molecule kinetic energies in the liquid phase would be slightly lower than in the solid phase. Moreover, the small change of nuclear kinetic energy across melting can be used to simplify the calculation of neutron-related environmental dose in complex locations, such as high altitude or polar neutron radiation research stations where liquid water and ice are both present: for neutron energies between hundreds of meV and tens of keV, the total scattering cross section per molecule in the two phases can be considered the same, with the macroscopic cross section only depending upon the density changes of water near the melting point. [ABSTRACT FROM AUTHOR]

Fe, Cr, and W are important elements in the alloys of in-reactor materials and operate in high-temperature environments with thermal expansion. Their tritium-impeding abilities are crucial to the radiation safety of various nuclear reactors. In this study, first-principles density functional theory is combined with quasi-harmonic approximation to evaluate factors that can affect the interstitial formation energy and diffusion coefficient of hydrogen isotopes in body-centered cubic (BCC) Fe, Cr, and W, including thermal expansion, metal host lattice vibrations, phonon density-of-states (pDOS) coupling diffusing atoms, and isotope effects. Calculation results indicate that the interstitial formation energy decreases as lattice expansion increases, whereas the jump barriers remain almost constant. Thermal expansion, host lattice vibration, and pDOS coupling minimally affect the diffusion coefficients of hydrogen isotopes in Fe, Cr, and W. The diffusion coefficient ratios between hydrogen isotopes are higher than the inverse ratio of the square root of the isotope mass at low temperatures. However, they decrease to the inverse ratio of the square root of the isotope mass at temperatures exceeding 800 K. This study comprehensively investigates factors that affect the diffusion coefficients of hydrogen isotopes in BCC Fe, Cr, and W, thus providing a firm theoretical foundation for predicting the diffusion coefficients of tritium at different temperatures using protium/deuterium diffusion coefficients. [ABSTRACT FROM AUTHOR]

This article investigates the under-explored potential of utilizing a thin stainless-steel diaphragm coupled with a quartz tuning fork sensor for liquid depth measurements. The focus is on monitoring molten salt fluid levels in nuclear reactors and concentrated solar power systems. Addressing a literature gap, the research explores cantilever-type configurations of a double-ended quartz tuning fork resonator, with a no-load resonance frequency of 17.37 kHz, on thin stainless-steel diaphragms for fluid depth measurement at room temperature. As the fluid depth increases, hydro-static pressure acting on a 20 μ m diaphragm causes deflection, bending a tuning fork. The resulting change in resonance frequency correlates with fluid depth. Experimental setups assess the tuning fork's sensitivity to strain and bending, revealing strain sensitivity of 7.83 Hz/ μ strain (450.78 ppm/ μ strain) and bending sensitivity of 0.09 Hz/ μ m (5.18 ppm/ μ m). The pressure sensor assembly, tested in a water tank, exhibits a sensitivity of − 0.28 Hz/mm (− 16.12 ppm/mm) in a single cantilever-type configuration. Despite a limited linear range, it effectively measures water depth changes as small as 0.7 mm. Exploring a double cantilever-type configuration yields a sensitivity of 0.07 Hz/mm (4.03 ppm/mm) with a broader linear range. The article discusses the reasons for opposite sensitivity and highlights the advantages of each configuration. Beyond molten salt level monitoring, the technology's applications may extend to fluid depth and pressure measurements in industrial and domestic settings. [ABSTRACT FROM AUTHOR]

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation. [ABSTRACT FROM AUTHOR]

In the 1999 Kargil War, India defended its territory from a Pakistani incursion but--in a departure from its historical behavior and standing war plans--chose not to expand the war with counter-attacks into Pakistan. Many observers attribute this restraint to nuclear deterrence, since India and Pakistan had become declared nuclear powers just a year earlier. In fact, India's restraint was rooted not in deterrence, but specific strategic conditions. Those conditions no longer apply--and in a future conflict India may be encouraged to take especially risky and escalatory wartime action, which would pose an unprecedented test for nuclear deterrence. [ABSTRACT FROM AUTHOR]

Scientists in Nazi Germany attempted but failed to build an atomic bomb or working nuclear reactor during the years of the Second World War, 1939-1945. Can philosophical meaning be drawn from this history? The history of the Nazi scientists and nuclear fission raises important issues concerning the alleged value-neutrality of scientific research and technological development. We can also ask whether there is an ethical dimension to the failure of the Nazi bomb project. Did the scientists who worked in the Third Reich make a moral decision not to complete the development of a working nuclear reactor and/or a usable nuclear explosive device? In this essay, I argue that Werner Heisenberg and other Nazi scientists fell victim to a value-neutral conception of science that led them to make unethical choices in support of the Nazi regime. The story of the Nazi bomb project exemplifies a more general philosophical point: if scientists fail to understand the social, political, and cultural context of their research they will be vulnerable to ethically wrong decisions. [ABSTRACT FROM AUTHOR]

ABSTRACT Dissimilar metal welds are utilized in the energy industry to connect two materials with different material characteristics. In the case of nuclear power plants, the connected materials tend to be low‐alloyed steel and high‐alloyed material. Despite different material and corrosion properties, under the proper environmental conditions, the used construction materials and weld metals are protected either by a passive layer or by a high‐temperature oxide. Although dissimilar metal welds are used in both primary and secondary circuits, the most frequently documented damage is in the secondary circuit, where, in addition to material heterogeneities, local environmental heterogeneities may form. For dissimilar metal welds in WWER nuclear power plants, a water–water energetic reactor, a subtype of pressure water reactor, we note two main types of attack near the dissimilar fusion boundary: on the carbon steel side or on the high‐alloyed weld metal side (X10CrNiMoN16‐25‐6). The possible causes of the latter “atypical” corrosion attack are debated and can be generalized as a consequence of change of the grain boundary condition. [ABSTRACT FROM AUTHOR]

Recent clinical success with metal-based radiopharmaceuticals has sparked an interest in the potential of these drugs for personalized medicine. Although often overlooked, the success and global impact of nuclear medicine is contingent upon the purity and availability of medical isotopes, commonly referred to as radiometals. For nuclear medicine to reach its true potential and change patient lives, novel production and purification techniques that increase inventory of radiometals are desperately needed. This tutorial review serves as a resource for those both new and experienced in nuclear medicine by providing a detailed explanation of the foundations for the production and purification of radiometals, stemming from nuclear physics, analytical chemistry, and so many other fields, all in one document. The fundamental science behind targetry, particle accelerators, nuclear reactors, nuclear reactions, and radiochemical separation are presented in the context of the field. Finally, a summary of the latest breakthroughs and a critical discussion of the threats and future potential of the most utilized radiometals is also included. With greater understanding of the fundamentals, fellow scientists will be able to better interpret the literature, identify knowledge gaps or problems and ultimately invent new production and purification pathways to increase the global availability of medical isotopes. [ABSTRACT FROM AUTHOR]

Conventional fluids used in fission‐based water‐cooled nuclear reactors have lower heat transfer coefficients (HTCs) and thermal conductivity, which has led researchers to explore high‐performance fluids that can enhance heat transfer in routine operation and prevent core meltdown in the case of accidents. It is important to investigate a wide range of fluids that can help designers improve thermal hydraulic characteristics, such as HTC, critical heat flux, and minimum departure from nucleate boiling ratio (MDNBR). In this study, the effectiveness of nanofluids in enhancing heat transfer parameters, including thermal conductivity and heat capacity, was investigated. Four different nanofluids (Al2O3–H2O, ZrO2–H2O, Ag–H2O, and Si–H2O) with pure water as the primary coolant in an HPR‐1000 nuclear reactor were compared using computational methods. Due to computational limitations, only the flow channel among four fuel rods with the highest power density in the core was simulated using Eulerian computational fluid dynamics. The results of this study show that silver water (Ag–H2O) nanofluid outperformed other nanofluids and pure water. It had a higher average HTC and MDNBR, with a 67.15 % and 45.23 % improvement, respectively, compared to pure water. The fuel rod wall temperature was also reduced by 28.5 K with Ag–H2O compared to water. Comparison of current simulated results with literature data shows a good agreement. [ABSTRACT FROM AUTHOR]

AlN/Er 2 O 3 multi-layer coatings, as a novel hydrogen isotope permeation barrier for nuclear reactor and hydrogen energy field, were prepared by magnetron sputtering coupled with metal-organic decomposition on the surface of reduced activation ferrite/martensitic steel. The thickness of the Er 2 O 3 layers was adjusted by spin coating several times, the adhesion of the layers was tested with a micron scratch tester, and the microstructure and surface composition of the layers were analyzed by TEM and XPS. The results indicate the absence of evident crack holes on the coating surface, uniform distribution of elements without segregation, and satisfactory densification. In comparison with the individual Er 2 O 3 coating, the multi-layer coating incorporating AlN as the interlayer exhibits enhanced adhesion strength, which could effectively mitigate the thermal mismatch phenomenon of Er 2 O 3 under high temperature conditions. Electrochemical hydrogen permeation and high temperature gas pressure driven deuterium permeation tests demonstrated that the permeability resistance of the AlN/Er 2 O 3 coating does not linearly increase with its thickness. The hydrogen permeation reduced factor of the coating reaches its maximum value when the thickness of the Er 2 O 3 coating is 1500 nm. [ABSTRACT FROM AUTHOR]

Accurate assessment of uncertainties in cross-section data is crucial for reliable nuclear reactor simulations and safety analyses. In this study, we focus on the interpolation procedure of the partial derivatives (PD) cross-section model used to evaluate nodal parameters from pregenerated multigroup libraries. Our primary objective is to develop a systematic methodology that enables prediction of the incurred errors in the cross-section model, leading to the development of optimal case matrices, more efficient cross-section models, and informed case matrix construction for reactor types lacking substantial data and experience. We make progress toward this objective through a rigorous analytic error analysis enabled by the derivation of error expressions and bounds for the PD model based on the discovery that the method is a form of Lagrange interpolation. Our investigations reveal distinct outcomes depending on the chosen cross-section functionalizations, achieved by identifying the sources of error. These error sources are found to include interpolation error, which is always present, and model form error, which is a property of the supplied case matrix. We show that simply increasing grid refinement through the addition of branches may not always lead to decreased cross-section errors, particularly in cases where the model form error predominantly contributes to the total error. We present numerical results and verification in a companion paper, showing these different error characteristics for various cross-section functionalizations. Although applied to current light water reactor environments, our methodology offers a means for advanced reactor analysts to develop case matrices with quantified error levels, advancing the goal of a general methodology for robust two-step reactor analysis. Future work includes exploring different lattice types and functionalizations, extending reactivity bounds to multilattice problems, and investigating historical effects within the macroscopic depletion model. [ABSTRACT FROM AUTHOR]

Due to the multiple physics involved and their mutual and complex interactions, nuclear engineers and researchers are constantly working on developing highly accurate Multi-Physics models, focusing in particular on the core coupling of Neutronics and Thermal-Hydraulics. Nevertheless, the development of accurate and stable models remains a challenging task despite the advancements in computational hardware and software. This work investigates the possibility of combining the available mathematical model with data collected on physical systems, with a two-fold goal: improvement of the performance of the former from the computational point of view without sacrificing accuracy and performing model bias correction with the knowledge coming in real-time from the system. In particular, two Data-Driven Reduced Order Modelling techniques, the Generalised Empirical Interpolation Method and the Parametrised-Background Data-Weak formulation, are applied to literature benchmark nuclear case studies, as they were observed to be quite well suited for the chosen cases. The main goal of this work is to assess the possibility of using external data to perform model bias correction: starting from a purposefully less accurate, but computationally cheaper base model, then using high-fidelity data to update and correct the model efficiently. Indeed, the numerical results obtained in this paper are promising, confirming the feasibility of this approach to develop computationally cheap and accurate multi-physics models; furthermore, investigation of Data-Driven Reduced Order Modelling approach to nuclear industrial cases, in the context of model bias correction, is foreseen. • State estimation using Data-Driven Reduced Order Modelling. • Solution of low-fidelity Multi-Physics models embedding Reduced Order Modelling approaches. • Update/correction of Multi-Physics models using higher-fidelity data. • Enhancing Multi-Physics modelling of nuclear reactors with data from different fidelity levels. [ABSTRACT FROM AUTHOR]

Fusion–fission hybrid reactors are concepts of subcritical reactors based on the coupling of fusion and fission devices. In this case, the fusion reactor would work as an external neutron supplier for the fission core of the machine. Such systems could, in principle, operate as multi-purpose machines, such as energy generators, breeders and waste burners. The large availability of fusion and fission technologies makes the choice of devices to couple quite chaotic. In fact, most of the concepts proposed in the literature are based on attempts without real optimization. The purpose of this paper is to propose a parameter which could provide practical information regarding the choice or the design of the fusion system of an FFHR. An engineering approach based on the estimation of the energy efficiency of FFHRs was used. An evaluation of the parameter and some of its possible practical applications are shown. Obtained results indicate that, from a geometrical point of view, compact machines would need lower Q-values to reach high neutron source performance. [ABSTRACT FROM AUTHOR]

The present study focuses on the recovery of waste heat in an autonomous safety system designed for advanced nuclear reactors. The system primarily relies on passive safety condensers, which are increasingly integrated into the design of advanced Pressurized Water Reactors (PWRs). These condensers are typically immersed in large water tanks that serve as heat sinks and are placed at sufficient heights to ensure natural circulation. Such a heat removal system can operate for an extended period, depending on the size of the tank. This research is driven by the potential to recover part of the energy stored in the boiling water volume, using it as a heat source for an Organic Rankine Cycle (ORC) system via an immersed heat exchanger. The electricity generated by the ORC engine can be used to power the system components, thereby making it self-sufficient. In particular, a pump replenishes the water tank, ensuring core cooling for a duration no longer limited by the water volume in the tank. An experimental test setup, including a boiling water pool and an ORC engine with an electrical output of approximately several hundred watts, along with an immersed evaporator, was constructed at CEA (Grenoble, France). Several test campaigns were conducted on the experimental test bench, exploring different configurations: two distinct ORC working fluids, cold source temperature variation effects, and relative positioning of the submerged evaporator and heat source within the water tank impact. These tests demonstrated the reliability of the system. The results were also used to validate both the ORC condenser and evaporator models. This article presents this innovative system, which has recently been patented. Moreover, to the best of our knowledge, the investigated configuration of an ORC that includes an immersed evaporator is original. [ABSTRACT FROM AUTHOR]

In developing materials for the nuclear industry, it is crucial to enhance both alloy composition and processing methods. This study focuses on investigations of applying radial-shear rolling (RSR) to a Zr-1%Nb alloy ingot, aiming to refine its microstructure and improve its properties for nuclear applications. This method, with complex vortex metal flow inside of a casted workpiece, has not been previously tested for processing zirconium ingots, so experimental verification of its applicability is of scientific interest. The 30 mm diameter ingot, produced by vacuum induction melting, was initially rolled to 20 mm at 800 °C to eliminate defects and refine the cast structure. A second rolling stage reduced the diameter to 13 mm at 530 °C, resulting in an ultrafine-grained structure. The RSR method effectively combines structural refinement and defect healing within fewer cycles, making it suitable for producing components for nuclear reactors. This approach demonstrates a potential reduction in traditional processing steps, providing a more efficient route for preparing high-quality materials for nuclear applications. [ABSTRACT FROM AUTHOR]

The deployment of gelcasted MAX phase ceramic components in nuclear reactors is critical due to their exceptional performance. This study explores the corrosion behaviors of three gelcasted MAX phase ceramics (Ti 3 AlC 2 , Ti 2 AlC, and Ti 3 SiC 2) in liquid lead-bismuth eutectic (LBE) at 600 °C under low oxygen conditions. Unique corrosion layers formed on the samples: Ti₃AlC₂ developed an Al 2 O 3 inner layer and TiO 2 outer layer, Ti₂AlC showed TiO 2 grains coated with FeCr 2 O 4 spinel, and Ti 3 SiC 2 exhibited a composite layer of TiO 2 , Fe 3 O 4 , SiO 2 , PbO, and TiO 2. LBE caused slight matrix enrichment in Ti₃AlC₂ and Ti₂AlC samples. Corrosion resistance was ranked as Ti 3 SiC 2 > Ti 3 AlC 2 > Ti₂AlC. Additionally, a coupled corrosion phenomenon resembling electrolytic electrodeposition between MAX phase ceramics and stainless steel was observed in contaminated LBE. This research provides novel insights into the corrosion mechanisms of MAX phases, paving the way for their application in advanced nuclear systems. [ABSTRACT FROM AUTHOR]

A heated junction thermocouple (HJTC) is a device designed to monitor the water level of nuclear reactors to prevent explosions due to overheating. This device consists of heated and non-heated thermocouples. When heat is applied, the temperature rise in HJTC varies depending on whether the device is submerged in water or exposed to air. In this study, we analyzed the thermal characteristics of HJTC through 3D COMSOL simulation. The temperature profile of the cross-section was extracted based on three design parameters: applied heat, heating length of the wound coil, and diameter of the HJTC. The temperature difference between the two thermocouples was more pronounced upon exposure to air. Meanwhile, high applied heat, short coil heating length, and small diameter of the HJTC contributed to a more pronounced temperature difference based on the surrounding environment. These results can aid in the appropriate HJTC design to achieve the desired detection sensitivity. [ABSTRACT FROM AUTHOR]

The McMaster nuclear reactor (MNR) is an important research facility that not only provides researchers with neutrons for fundamental science, but also supplies the medical industry with isotopes used for cancer treatment. To ensure the safety and performance of the MNR, modeling of the thermal hydraulics during nominal and accidental conditions is required. For such a task, system codes are customarily used. While system codes can assess the safety aspects of complex thermal–hydraulic systems, the question arises whether such systems can be modeled appropriately in a three-dimensional manner, such as computation fluid dynamics (CFD), while capturing the thermal mixing of the coolant. Modeling the thermal hydraulics of nuclear research reactors using CFD is relatively new. Therefore, validating such methods against experimental data is of utmost importance. The validation of CFD is the main focus of this work. More specifically, the influence of the loss of heat sink on the pool temperature is assessed using the CFD code Ansys CFX. The validation basis was provided by an experiment performed at the MNR in March 2023. In this experiment, the secondary heat removal system was intentionally shut down, and the pool temperature was measured in a few positions. The results obtained by modeling the loss of heat sink in the MNR using Ansys CFX agree well with the experiment. [ABSTRACT FROM AUTHOR]

Slow‐cooled CF8M duplex stainless steel is used for critical parts of the primary coolant pipes of nuclear reactors. This steel can endure severe service conditions, but it tends to become more brittle upon very long‐term aging (tens of years). Therefore, it is essential to understand its specific microstructure and temporal evolution. As revealed by electron backscatter diffraction (EBSD) analyses, the microstructure consists of millimetre‐scale ferritic grains within which austenite lath packets have grown with preferred crystallographic orientations concerning the parent ferritic phase far from the ferrite grain boundaries. In these lath packets where the austenite phase is nucleated, the lath morphology and crystal orientation accommodate the two ferrite orientations. Globally, the Pitsch orientation relationship appears to display the best agreement with the experimental data compared with other classical relationships. The austenite lath packets are parallel plate‐shaped laths, characterized by their normal n. A novel methodology is introduced to elucidate the expected relationship between n and the crystallographic orientation given the coarse interfaces, even though n is only partly known from the observation surface, in contrast to the 3D crystal orientations measured by EBSD. The distribution of retrieved normals n is shown to be concentrated over a set of discrete orientations. Assuming that the ferrite and austenite obey the Pitsch orientation relationship, the determined lath normals are close to an invariant direction of the parent phase given by the same orientation relationship. [ABSTRACT FROM AUTHOR]

The technology of nuclear reactors is evolving rapidly, driven by the pursuit of more powerful and efficient systems. The Small Lead-based Reactor (SLR) represents an advanced nuclear reactor design that holds great promise for delivering enhanced power and efficiency. In the context of the SLR's fuel rod, the high-speed coolant flow within the reactor can induce vibration, potentially causing fretting wear and damage to the cladding. This study utilized the Burgreen correlation to establish an equivalence relationship between water and Lead-Bismuth Eutectic (LBE). An experiment was then conducted to simulate axial flow-induced vibration (FIV) of a simply supported fuel rod, employing an equivalent water loop. Flow-induced vibration characteristics in both the time and frequency domains were investigated at different flow speeds. The experimental results revealed a positive correlation between flow velocity and amplitude. The fuel rod exhibited low-frequency vibrations with a random pattern, registering frequencies around 14 Hz. These experimental findings can be leveraged to enhance the accuracy of numerical simulations for axial flow-induced vibration (FIV) of the fuel rod. Subsequently, the revised numerical model was applied to simulate FIV in the LBE environment using computational fluid dynamics (CFD), and the numerical results were found to be in agreement with the experimental data. [ABSTRACT FROM AUTHOR]

The electrochemical corrosion potential (ECP) data of nickel-base alloy 82 weld metal (Alloy 82) in both O2 and H2O2 environments were measured. Although Alloy 82 is used at various welds in boiling water reactors (BWRs), the ECP data of Alloy 82 under BWR conditions are limited. The ECP data measured in this study revealed that Alloy 82 had higher ECPs than 304 and 316 L stainless steels and nickel-base alloy 182 weld metal (Alloy182) at low oxidant concentrations in both O2 and H2O2 environments. The ECPs of Alloy 82 calculated with an analytical ECP model that was developed in this study agreed with measured values in the cases that it was acceptable for the model to have an error range of ± 0.1 V in the tested range of O2 and H2O2 concentrations. The possible reason for the ECPs of Alloy 82 to be higher than those other materials at low oxidant concentrations is given as follows: the anodic polarization curve of Alloy 82 exhibits a lower current density at low potentials because of its higher Ni and Cr contents. The developed ECP model validated by the ECP data will contribute to evaluations of the effectiveness of mitigation technologies for stress corrosion cracking environments. [ABSTRACT FROM AUTHOR]

Stainless-steel is extensively utilized in the key structural components of the main equipment in the nuclear island of pressurized water reactor nuclear power plants. The operational experience of nuclear power plants demonstrates that stress corrosion is one of the significant factors influencing the long-term safe operation of stainless steel in the high-temperature water of pressurized water reactor nuclear power plants. This study is based on the stress corrosion crack growth rate data of 316SS and 304SS stainless steel in the simulated primary water environment of pressurized water reactor nuclear power plants. Data mining and modeling were conducted using multiple machine learning algorithms, including Random Forest (RF), eXtreme Gradient Boosting (XGBoost), Support Vector Regression (SVR), and Gaussian Process Regression (GPR), and the Sharpley Additive explanation (SHAP) method was employed to analyze the interpretability of the model. The results indicate that the stress corrosion crack growth rate prediction model based on XGBoost outperforms other models in all assessment indicators. Compared with empirical equations, XGBoost exhibits high flexibility and excellent data-driven learning capabilities. In the test set, 90% of the prediction errors are within the range of experimental values, with the maximum error multiple being 2.5, which significantly improves the prediction accuracy. Moreover, the distribution of SHAP values is consistent with the theoretical study of the stress corrosion behavior of stainless-steel, effectively reflecting the impact of cold working, temperature, and stress intensity factor on the stress corrosion crack growth rate, thereby proving the reliability of the model's prediction results. The achievements of this study hold significant reference value and application prospects for the prediction of the stress corrosion behavior of stainless-steel in a high-temperature and high-pressure water environment of pressurized water reactor nuclear power plants. [ABSTRACT FROM AUTHOR]

AbstractThe aging process or flow-induced vibration of reactor cores may lead to increased mechanical vibrations, affecting the reliability of in-core sensors and necessitating a robust solution for robust field reconstruction. This work tackles the challenges of reconstructing multiphysics fields from sparse and movable measurements by introducing an advanced framework that integrates various machine learning models with Voronoi tessellation. Our approach, building upon the Voronoi tessellation-assisted Convolutional Neural Network (VCNN), expands the capabilities to include a wider array of neural network architectures such as Convolutional Neural Networks (CNNs), Fourier Neural Operator (FNO), Dilated ResNet Encode-Process-Decode (DilResNet), Dilated Convolution Neural Operator (DCNO), Galerkin Transformer (GT), U-shaped Neural Operator (UNO), and Multiwavelet-based Operator (MWT). The effectiveness of these models is evaluated and validated through numerical tests based on the International Atomic Energy Agency benchmark, particularly noting average relative errors below 5% and 10% in the $${L_2}$$L2 norm and $${L^\infty }$$L∞ norm, respectively, within a 5-cm amplitude around sensor nominal locations. The developed software toolkit encapsulates these architectures, providing a versatile option for nuclear engineers to reconstruct different types of physical fields efficiently. [ABSTRACT FROM AUTHOR]

This study investigates the impact of hybrid nanoparticles on the temperature of nuclear reactor coolant, with a focus on graphene nanoplatelet (GNP)‐based hybrid nanoparticles. Sixteen different hybrid nanofluids were analyzed, and their performance was compared with a standard water‐based coolant. The criticality values were obtained through MCNP modeling, revealing that higher nanoparticle ratios led to increased criticality, with the highest value of 1.3239 observed in GNP‐Fe3O4 + Al2O3 nanofluids (0.05 wt%) and the lowest value of 1.2935 in GNP–Fe3O4 + SiO2 nanofluids (0.001 wt%). Temperature variations showed that increasing nanoparticle concentrations resulted in slightly higher temperatures, with a maximum of 611.97 K for 0.05 vol.% GNP nanoparticles. Additionally, the departure from nucleate boiling ratio values were consistently above the safety threshold of 2.08, with the lowest value of 3.657 for GNP–Fe3O4 + SiO2 nanofluids (0.05 vol.%). These findings suggest that hybrid nanofluids, particularly those with higher nanoparticle ratios, can enhance the thermal performance and safety margins of nuclear reactor coolants, offering a promising avenue for future research and application. [ABSTRACT FROM AUTHOR]

This investigation has significant applications in the fields of mechanical, industrial, and biomedical engineering, biosciences, and technology, particularly in areas such as blood pumping, drug delivery, glass-fiber production, paper production, and nuclear reactors. This study explores the numerical analysis of a mathematical model of entropy generation in convective magnetohydrodynamic (MHD) flow of Williamson nanofluid model over a stretching sheet with Cattaneo–Christov heat flux and suction/injection. Furthermore, heat generation, viscous dissipation, Joule heating, radiation, and chemical reaction are considered. The appropriate transformations are used to transform the nonlinear partial differential equations (PDEs) into nonlinear ordinary differential equations (ODEs). The transformed equations are solved numerically using bvp5c MATLAB package. The effects of the involved physical parameters on the flow quantities and the entropy generation are presented and discussed in detail with figures and tables. It is observed that the thermal field is enhanced by increasing the Eckert number, the Joule heating, and the thermal relaxation parameters. Also, the concentration field is observed to be a decreasing function of the augmented chemical reaction parameter. Further, the increasing magnetic field and the Williamson parameter led to an increase in the skin friction coefficient. Moreover, the entropy generation increases due to an increase in the diffusion parameter and the Brinkman number. [ABSTRACT FROM AUTHOR]

This article focuses on apartheid South Africa’s nuclear sector and the regime’s attempts to cooperate with overseas energy companies to provide financial investments and crucial technologies. It highlights the connectedness of the South Africans in the global nuclear marketplace and their ability to secure technical support from Western states during the Cold War between 1953 and 1976. The article analyses the parallel negotiations with French and German firms to engage with the regime in sharing of sensitive knowledge and bargaining lucrative contracts. Using newly discovered archival records it is shown that French-German competition was at a boiling point at least twice in the 1970s. Moreover, the regime in Pretoria managed to garner enough support in the nuclear field to further the growth of its domestic industry, ultimately being capable of enriching uranium and obtaining a turnkey nuclear power reactor. The white minority government in Pretoria however failed to position itself as an important uranium supplier on a global scale, because of sanctions targeting domestic racial apartheid policies and a more robust international non-proliferation regime towards the 1980s. [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) studs are key components of nuclear reactors, and their connection with flange ensures the sealing of the RPV under high-pressure and high-temperature conditions. In the present work, the external threads of the RPV stud were prepared by triaxial rolling, and the texture evolution of the external thread root material of an RPV stud was predicted by finite element analysis coupled with viscoplastic self-consistent simulations. The microstructure of the external thread root material of RPV stud was characterized by scanning electron microscope and electron back-scattered diffraction. The installation characteristics of the turned and rolled parts of the RPV stud were tested by installation and pretightening tests. It was found that the dynamic recrystallization at the external thread root formed ultrafine tempered sorbite grains, high-angle grain boundaries (47%), and strong {111} <110> and {111} <112> textures. In the installation and pretightening test, the residual elongation of rolled parts was reduced by 6% under the same loading pressure. The triaxial rolling process distributed the microstructure of the external thread root of the RPV stud in a gradient manner, resulting in improved stud installation characteristics. [ABSTRACT FROM AUTHOR]

Introduction Military members and first responders may, at moment's notice, be asked to assist in incidents that may result in radiation exposure such as Operation Tomadachi in which the U.S. Navy provided significant relief for the Fukushima Daiichi Nuclear Reactor accident in Japan after an earthquake and tsunami in 2011. We are also currently facing potential threats from nuclear power plants in the Ukraine should a power disruption to a nuclear plant interfere with cooling or other safety measures. Exposure to high doses of radiation results in acute radiation syndrome (ARS) characterized by symptoms arising from hematopoietic, gastrointestinal, and neurovascular injuries. Although there are mitigators FDA approved to treat ARS, there are currently no FDA-approved prophylactic medical interventions to help protect persons who may need to respond to radiation emergencies. There is strong evidence that manganese (Mn) has radiation protective efficacy as a promising prophylactic countermeasure. Materials and Methods All animal procedures were approved by the Institutional Animal Care and Use Committee. Male and female B6D2F1J mice, 10 to 11 weeks old, were used for neurotoxicity studies and temporal effects of Mn. Four groups were evaluated: (1) vehicle injection, (2) dose of 4.5 mg/kg for 3 days, (3) dose of 13.5 mg/kg, and (4) sham. Irradiated mice were exposed to 9.5 Gy whole body Co60 γ-radiation. MRI was performed with a high dose of manganese chloride (MnCl2) (150 mg/kg) to assess the distribution of the MnCl2. Results The mice have promising survival curves (highest survival—13.5 mg/kg dose over 3 days of MnCl2 at 80% [87% female, 73% male] P  = 0.0004). The complete blood count (CBC) results demonstrated a typical hematopoietic response in all of the irradiated groups, followed by mildly accelerated recovery by day 28 in the treated groups. No difference between groups was measured by Rota Rod, DigiGait, and Y-maze. Histologic evaluation of the bone marrow sections in the group given 13.5 mg/kg dose over 3 days had the best return to cellularity at 80%. MRI showed a systemic distribution of MnCl2. Discussion The preliminary data suggest that a dose of 13.5 mg/kg of MnCl2 given over 3 days prior to exposure of radiation may have a protective benefit while not exhibiting the neurobehavioral problems. A countermeasure that can prophylactically protect emergency personnel entering an area contaminated with high levels of radiation is needed, especially in light that nuclear accidents are a continued global threat. There is a need for a protective agent with easy long-term storage, easy to transport, easy to administer, and low cost. Histologic evaluation supports the promising effect of MnCl2 in protecting tissue, especially the bone marrow using the dose given over 3 days (4.5 mg/kg per day) of MnCl2. Conclusions Initial experiments show that MnCl2 is a promising safe and effective prophylactic countermeasure against ARS. MRI data support the systemic distribution of MnCl2 which is needed in order to protect multiple tissues in the body. The pathology data in bone marrow and the brain support faster recovery from radiation exposure in the treated animals and decreased organ damage. [ABSTRACT FROM AUTHOR]

Photodetectors used in future high-energy physics experiments will need to keep sufficient performance during a few years of data-taking despite radiation load, which, for example, in the planned upgrade of the Ring Imaging Cherenkov detectors in Large Hadron Collider beauty (LHCb) experiment is estimated at about 10 13 1-MeV neutron equivalent per cm 2 (neq/cm 2 ). Silicon photomultipliers (SiPMs) are considered as candidates for photodetectors for this application. However, their sensitivity to neutron irradiation may seriously compromise their operation, with the increase in dark count rates being the primary limitation after the irradiation. In this work, 1 × 1 mm 2 15 μ m pitch NUV-HD-RH silicon photomultipliers developed by the Fondazione Bruno Kessler were characterized before and after the irradiation. In total, 5 SiPMs were irradiated at the Jožef Stefan Institute TRIGA nuclear reactor with different fluences from 10 9 neq/cm 2 up to 10 13 neq/cm 2 . The SiPMs were also annealed at high temperatures and re-characterized after the annealing. For the SiPM characterization in all the cases, current–voltage (I-V curve) measurements, dark count rate measurements, and waveform analysis, including single photon time resolution, were carried out at different controlled temperature steps from room temperature down to the liquid nitrogen temperature. While cooling to − 20 ∘ C was enough for the SiPM irradiated at 10 9 neq/cm 2 to recover the ability to resolve single photons, by 10 13 neq/cm 2 , cooling to liquid nitrogen temperature was necessary. With sufficient cooling, the measured single photon time resolution was around 90 ps FWHM for all irradiation levels. [ABSTRACT FROM AUTHOR]

The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data. [ABSTRACT FROM AUTHOR]

Disruptions in tokamak nuclear reactors, where plasma confinement is suddenly lost, pose a serious threat to the reactor and its components. Classifying discharges as disruptive or non-disruptive is crucial for effective plasma operation and advanced prediction. Traditional disruption identification systems often struggle with noise, variability, and limited adaptability. To address these challenges, we propose an enhanced stacking generalization model called the "Double-Phase Stacking Technique" integrated with Pool-based Active Learning (DPST-PAL) for designing a robust classifier with minimal labor cost. This innovative approach improves classification accuracy and reliability using advanced data analysis techniques. We trained the DPST-PAL model on 162 diagnostic shots from the Aditya dataset, achieving a high accuracy of 98% and an F1-score of 0.99, surpassing conventional methods. Subsequently, the deep 1D convolutional predictor model is implemented and trained using the classified shots obtained from the DPST-PAL model to validate the reliability of the dataset, which is tested on 47 distinct shots. This model accurately predicts the disruptions 7–13 ms in advance with 93.6% accuracy and exhibited no premature alarms or misclassifications for our experimental shots. [ABSTRACT FROM AUTHOR]

The natural circulation loop is crucial for the safe and stable operation of nuclear reactors and other applications. Traditional numerical algorithms, based on the Boussinesq approximation, have limitations when dealing with large temperature differences and density disparity, and they do not fully address fluid compressibility. This paper adopts the decoupled and stabilized lattice Boltzmann method (DSLBM) with a non-Boussinesq algorithm to study the natural circulation loop. The DSLBM provides a detailed flow description under large temperature and density differences, incorporating the pseudopotential multiphase model, temperature equation, and state equation, without relying on assumptions. The study examines the loop's performance under various temperature differences, central height differences, and heating source lengths, focusing on mass flow rate, driving head, and heating power. It reveals the energy performance, flow characteristics, and heat transfer properties of the loop, highlighting the physical mechanisms involved. Comparison with the empirical formulation of the incompressible equation from the theoretical aspect shows that when the temperature difference coefficient is lower than 0.15, the two methods are not much different from each other. When the temperature difference coefficient reaches 0.2, 0.3, and 0.4, the difference between the two methods is 9.47%, 19.11%, and 42.64%, respectively. Consequently, the Boussinesq approximation can be compensated by DSLBM, which proves the value of the application of the algorithm in exploiting the compressibility of fluids. The dimensionless fitting correlation with greater universality is obtained, which helps to predict the properties of the natural circulation loop with varying temperature differences, friction coefficients, and geometric structures. The research in this paper will lay the foundation for optimizing the system design of the natural circulation loop and improving energy utilization efficiency. [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]

To mitigate global warming, phasing out coal in the global energy system orderly and rapidly is an important near-term strategy. However, the majority of coal-fired plants in China have operated for less than 15 years. Accelerated coal power plant retirements would lead to substantial asset stranding. Coal-to-nuclear (C2N) technology offers a potential solution by replacing coal boilers in existing coal-fired plants with nuclear reactors. In this study, the G4-ECONS model was used to assess the economics of repowering a 600 MW supercritical coal-fired power plant with two 272 MWe high-temperature gas-cooled reactors. The timeline for the C2N project and the additional cost of dispatching electricity from the grid during retrofitting were discussed. Results showed that the C2N total capitalized costs are 19.4% (baseline estimate, USD 5297.6/kW) and 11.1% (conservative estimate, USD 5847.2/kW) lower than the greenfield project (USD 6576.5/kW), respectively. And C2N projects need to reduce LUEC by at least 20% to become competitive. This study can inform engineering design decisions leading to more precise and cost-effective C2N projects. [ABSTRACT FROM AUTHOR]

High-entropy alloys (HEA) are promising structural materials that will successfully resist high-temperature irradiation with helium ions and radiation-induced swelling in new generations of nuclear reactors. In this paper, changes in the elemental and phase composition, surface morphology, and structure of CoCrFeNi and CoCrFeMnNi HEAs irradiated with He2+ ions at a temperature of 700 °C were studied. Structural studies were mainly conducted using the X-ray diffraction method. The formation of a porous surface structure with many microchannels (open blisters) was observed. The average diameter of the blisters in CoCrFeMnNi is around 1.3 times smaller than in CoCrFeNi. It was shown that HEAs' elemental and phase compositions are stable under high-temperature irradiation. It was revealed that, in the region of the peak of implanted helium, high-temperature irradiation leads to the growth of tensile macrostresses in CoCrFeNi by 3.6 times and the formation of compressive macrostresses (−143 MPa) in CoCrFeMnNi; microstresses in the HEAs increase by 2.4 times; and the dislocation density value increases by 4.3 and 7.5 times for CoCrFeNi and CoCrFeMnNi, respectively. The formation of compressive macrostresses and a higher value of dislocation density indicate that the CoCrFeMnNi HEA tends to have greater radiation resistance compared to CoCrFeNi. [ABSTRACT FROM AUTHOR]

Digitally controlled reactors can optimize biological reactions and process control through a neural network system. This study reports on the design, fabrication, and automation of a laboratory‐scale anaerobic reactor for the management of agrifood byproducts and bioenergy recovery. The process described here can digitally control the operational parameters, which is beneficial for stable methane production. The proposed process comprises the digital measurement of temperature, pH, humidity, biogas volume, and methane composition by integrating the data in a processor module. The proposed automated reactor can assist significantly in controlling and monitoring the anaerobic digestion process, providing decision making during waste management and bioenergy recovery. A case study is described with the application of automated reactors in a pilot‐scale plant, operated with the flow of 8 m3 slaughterhouse wastewater per day and a biogas production of 10 m3 h−1. The automated pilot‐scale process presents many advantages, including a continuous mode of operation and a faster adaptation of the microorganisms to the substrate, improving biogas production. [ABSTRACT FROM AUTHOR]

In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant. [ABSTRACT FROM AUTHOR]

Plate-type fuel elements, widely used in high flux reactor (HFR), may experience collapse or vibration under high coolant flow rates. The study examined the behaviors of curved fuel plates under varying coolant velocities. The deformations of curved plates with varying curvature radii and flat-plate assemblies were computed to facilitate comparison. The fluid domain was simulated using Ansys Fluent, while the solid domain was analyzed using Ansys Mechanical. The study employed a partitioned, two-way, and explicit coupling strategy. The simulation results have shown that the critical velocity for assemblies of curved plates is 20 m/s, whereas for flat plates it is 6 m/s. As the curvature radius or width increases, the plates become increasingly susceptible to collapse. The critical velocity is associated with the nonlinearity between the maximum plate deflections and the square of flow velocities. It also involves plate deflections that extend from the leading edges to the entire length of the fuel at the critical velocity. Under axial flow conditions, the deflection directions of multiple curved plates exhibit regular patterns. The leading edges of the plates adjacent to the walls deflect towards the walls, while the adjacent plates deflect in opposite directions. [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.)

In view of the complexity and importance of nuclear research reactor (NRR) installations, it is imperative to uphold high standards of reliability and quality in the electricity being supplied to them. In this paper, the performance of low-voltage (LV) distribution feeders integrated with NRRs is improved in terms of reduced distribution loss, improved voltage profile, and reduced greenhouse gas (GHG) emissions by determining the optimal location and size of photovoltaic (PV) systems. In the second stage, the power quality of the feeder is optimized by reducing the total harmonic distortion (THD) by optimally allocating D-STATCOM units. In the third and fourth stages, the reliability and resilience aspects of the feeder are optimized using optimal network reconfiguration (ONR) and by integrating an energy storage system (ESS). To solve the non-linear complex optimization problems at all these stages, an efficient meta-heuristic Chernobyl disaster optimizer (CDO) is proposed. Simulations are performed on a modified IEEE 33-bus feeder considering the non-linear characteristics of NRRs, variability of the feeder loading profile, and PV variability. The study reveals that the proposed methodology can significantly improve the service requirements of NRRs for attaining sustainable research activities. [ABSTRACT FROM AUTHOR]

With a suitable half-life and abundant radiolabeling strategy, 225Ac has become one of the most promising radionuclides in the area of targeted alpha therapy. However, limited radionuclide supply is threatening the development of 225Ac related endoradiotherapy dramatically. As the parent nuclide of 225Ac, 229Th can be produced via 226Ra(3n, 2ß)229Th reaction in a nuclear reactor. However, related practice has not been conducted in large scale, since the nuclear reaction pathway for producing 229Th is complicated. In this work, the feasibility of producing 225Ac/229Th in a reactor was confirmed by systematic theoretical calculations, and a procedure that combines irradiation with separation process was proposed. The results show that 176 MBq of 229Th can be produced by irradiating 1.0 g of 226Ra with a neutron flux density of 1 × 1015 n cm-2 s1 for 90 days. This will generate 150 MBq of 225Ac monthly from a radionuclide generator, which is sufficient for the single treatment cycle of 200 patients each year considering the radioactivity loss in radiochemical separation, transfer and radiolabeling process. In addition, this irradiation process will also produce 37.8 GBq 227Ac for the preparation of 227Ac-227Th-223Ra generator. In general, the production of 225Ac by neutron irradiation of 226Ra in reactor is practicable and holds potential to alleviate the shortage of current supply of 225Ac. [ABSTRACT FROM AUTHOR]

The survivability of the domestic nuclear power industry depends on the cost-competitiveness of safe and secure nuclear power generation. Advanced reactor design concepts aim to have increased safety margins over traditional large light water reactors (LWRs). With increased safety margins comes the potential for a corresponding decrease in off-site risk to the general public from a hypothetical release of radioactivity due to sabotage or theft. Without sacrificing safety or security, advanced reactor designers may be able to achieve operational cost improvements over current LWRs in part by designing less burdensome physical protection systems (PPSs) and by replacing on-site response forces with off-site response forces. To accommodate these developments, the U.S. Nuclear Regulatory Commission is drafting new rulemakings for physical security when licensing through the current frameworks in 10 CFR 50 or 10 CFR 52 along with drafting an entirely new licensing framework: 10 CFR 53. A novel technology-inclusive consequence-informed methodology for the selection of the optimal licensing path for the design of PPSs at advanced fixed-site commercial nuclear power facilities is presented herein. This methodology proposes integrating security considerations at the beginning of a reactor facility design effort to streamline the licensing process. Off-site total effective dose equivalents at the exclusion area and low population zone boundaries were identified as the key metrics when determining a design's most appropriate licensing path that in turn affects the design requirements placed upon the PPS. Given these metrics, source-term generation of potential adversary-induced physics-based sabotage actions utilizing severe accident modeling software and off-site plume dispersal modeling were identified as appropriate for determining siting constraints, potential target sets for hypothetical sabotage events, and their subsequent off-site dose consequences. The methodology proposes using the consequence results from the sabotage modeling, in combination with desired cost-saving PPS characteristics, to help inform the licensing path selection. Once a licensing path is chosen, the methodology utilizes the Design and Evaluation Process Outline to evaluate an effective PPS following the licensing requirements placed on the facility. This paper also presents examples of hypothetical commercial nuclear power facilities with varying consequence levels and demonstrations of how to select the optimal licensing pathways for each. [ABSTRACT FROM AUTHOR]

On 25–26 March 2023, the U.S. National Council on Radiation Protection and Measurements (NCRP) held its 2024 annual meeting in Bethesda, Maryland, USA. The NCRP dates from 1929, and this meeting celebrated the 60th anniversary of receiving a U.S. Congressional Charter. For this annual meeting the NCRP felt it was essential to provide a briefing about advanced and small modular nuclear reactors (SMRs). The Journal of Radiological Protection is delighted to publish the following synopsis of material presented at the U.S. NCRP meeting. This synopsis is divided into five sections. The first section provides an overview of the whole meeting together with summaries of two context setting overview papers. The following four sessions of this synopsis are specific to advanced and small modular nuclear power reactors. The meeting also included keynote presentations by three of NCRP annual award recipients. The meeting topical areas were Technology Overview and Critical Issues. The individual papers laid the groundwork to understanding reactor technologies, terminology, and the fundamental concepts and processes for electrical generation. The perspectives of the U.S. Environmental Protection Agency and states, through the Conference of Radiation Control Program Directors were provided. The papers included a discussion of diverse topics including potential emergency preparedness considerations, radiological survey requirements, an evaluation of the future of nuclear power, the economics of reactors (both large and small), and the critical issues identified by the recent National Academies of Sciences' study on advanced reactors. The summary papers were developed to briefly document the major points and concepts presented during the oral papers presented at the 2024 NCRP Annual Meeting. The meeting heralded the dawn of a new era for commercial nuclear power. [ABSTRACT FROM AUTHOR]

Author(s): Arun Chougule (corresponding author) [1,2,3]; Gourav Kumar Jain [4] Author: The IAEA officers responsible for this publication were M. Carrara and E. Titovich of the Division of Human Health [...]

Nb is considered a promising candidate as a refractory element due to its high-temperature endurance, excellent thermal conductivity, and compatibility with liquid-metallic coolants in nuclear reactors. In the present study, radiation-based molecular dynamics numerical simulations were conducted in Σ 13, Σ 29, and Σ 85 symmetric tilt grain boundary models for pure Nb specimens. The stochastic high-energy collisions were modeled via large-scale atomic/molecular parallel simulator code to accurately investigate the radiation-induced defects generated in the order of picoseconds at the atomic level. The long-range embedded atom method potential and coulombic repulsive Ziegler–Biersack–Littmark potentials were smoothly overlaid for precise force-field interactions among Nb atoms. To investigate the ability to arrest the radiation-induced damage, the bi-crystal Nb specimens were irradiated at varying magnitudes of primary-knock-on atom (PKA) energies EPKA = 10 20, and 30 keV at temperature regimes 300, 600, and 900 K, respectively. The Frenkel pairs, complex linear defects, distribution of point defects as clusters, rate of defect annihilation, and temperature fluctuations within the displacement cascades of irradiated Nb specimens were comprehensively studied and reported. Here, the Nb-Σ 29 GB model survived with the lowest number of residual defects. Also, the recombination rate of the irradiated Nb specimens increases with the increase in temperature and PKA energy magnitude due to enhanced atomic mobility of the dislodged atoms. Hence, the bi-crystal Nb specimen can be favored for a radiation-tolerant material as structural components in next-generation reactors. [ABSTRACT FROM AUTHOR]

Recently declassified material and other information that has never before appeared in the public domain allow the authors to explain some of the workings of the Nuclear Emergency Support Team (NEST)—often one of the first units to respond whenever there is a nuclear incident, whether it involves a nuclear reactor or a nuclear weapon. Long the subject of mystique, NEST is often depicted on screen as a secretive government unit with highly specialized capabilities and harrowing missions. The reality is at once more mundane and more remarkable. Formed in the 1970s in response to a spate of nuclear blackmail attempts, NEST has been at the center of every major nuclear event from the accident at Three Mile Island to the disaster at Fukushima. Other operations, unknown to the public, are described here for perhaps the first time. Historical accounts provide a glimpse into the breadth of the organization's missions, from neutralizing terrorist nuclear devices to responding to nuclear reactor accidents. The diversity of NEST's missions and the uniqueness of its scientific capabilities set the unit apart as a national asset. [ABSTRACT FROM AUTHOR]

Zirconium alloys are used as a cladding material for fuel elements in nuclear reactors. In the case of severe accident conditions, the possible rapid oxidation of zirconium in steam or/and air may result in intense hydrogen generation and hydrogen–oxide mixture explosion. Advanced technologies for increasing the corrosion resistance of claddings are being investigated in two directions: (a) protective coatings on Zr alloys and (b) the use of new materials for claddings. Coatings with silicon may provide a more protective barrier than the ZrO2 films formed on an alloy cladding during nuclear plant operations. These coatings may also serve as a protective barrier during high-temperature accidents. The current work aimed at developing protective coatings with silicon on zirconium alloys. Multielemental Zr–Si–Cr coatings were formed on Zry-2 alloy using the physical vapor deposition (PVD) method. Long-term oxidation tests were carried out under the following conditions: 360°C/195 bar/63 days/water-simulating PWR water. Obtained results show the protective character of formed layers. The material in the form of silicon carbide grains covered with yttrium–aluminum garnet (SiC + YAG) was prepared using the sol–gel method. The formed powder is the main component for coating formation on Zr–1Nb alloy using the method of suspension plasma spraying (SPS).

Digital engineering and digital twins are increasingly being used in nuclear energy projects with important impacts. At Idaho National Laboratory, these approaches have been applied in a variety of nuclear energy research, development, and demonstration projects, with key lessons and evolutions occurring for each. In this paper, we describe the use of digital engineering and digital twins in the Versatile Test Reactor design, National Reactor Innovation Center test beds, and nonproliferation analysis of the AGN-201 reactor design. We share key lessons learned for these projects related to tool selection, adoption and training, and working with existing assets versus beginning at the design phase. We also share highlights of future potential uses of digital twins and digital engineering, including using artificial intelligence to perform repetitive design tasks and digital twins to move towards semiautonomous nuclear power plant operations. [ABSTRACT FROM AUTHOR]