A U + 5 wt.% Al powder (written U-5Al), manufactured by centrifugal atomization, was used in this study to assess the feasibility of densifying UAl x + Al powder mixtures by Spark Plasma Sintering (SPS) and obtaining a final composition with a minimum content of UAl 2 that meets specifications of targets for 99 Mo production. The microstructure of the U-5Al powder was studied first and mainly consists of a mixture of UAl 2 primary dendrites and α -U phase. Three SPS conditions differing in terms of UAl x /Al volume ratio and dwell time at 800 K were tested. The sintered discs were characterized by X-ray diffraction, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy. A very quick transformation of the U-5Al particles into a mixture of aluminides (UAl 2 + UAl 3 + UAl 4 , referred to UAl x ) was evidenced: only a few minutes were sufficient to transform the U-based particles into a mixture of aluminides containing about 80 wt.% of UAl 3 . According to these first tests, this new highly versatile manufacturing process could advantageously replace the conventional method of producing targets, subject to technical adjustments and parametric studies.
FeCrAl alloys are promising candidate for accident tolerant fuel (ATF) cladding in light water-cooled nuclear reactors (LWRs) due to their excellent high-temperature oxidation resistance and mechanical strength. This study investigates the effect of silicon (Si) additions (0–8.00 wt.%) on the neutronic performance and microstructural evolution of FeCrAl alloys through combined Monte Carlo simulations and experimental characterization. Neutronic behavior was evaluated using OpenMC on an AP1000 pressurized water reactor assembly model, while microstructural and mechanical properties were assessed via optical microscopy, SEM/EDS, XRD, and Vickers hardness testing. The results show that incorporating Si slightly improves the neutron economy of FeCrAl cladding by reducing Fe content, though the infinite multiplication factor ( k inf ) remains lower than that of Zr-4, limiting fuel discharge to ∼830 days compared to ∼1160 days for Zr-4. Experimentally, increasing Si content refined the average grain size from ∼498 μm (0 wt.% Si) to ∼201 μm (8.00 wt.% Si), accompanied by a linear increase in hardness from ∼280 HV to ∼520 HV. XRD analysis confirmed lattice parameter shifts consistent with Si incorporation, while SEM/EDS revealed generally uniform elemental distribution with localized Si segregation at higher concentrations. These findings highlight a trade-off: Si enhances microstructural stability and mechanical performance but only partially mitigates the neutronic penalty of FeCrAl. Optimizing Si content in conjunction with reduced cladding thickness is proposed as a pathway to balance neutronic efficiency with mechanical robustness for ATF deployment.
This paper presents the work performed in the subgroup 16 of the Working Party for Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. The main goal was to define two decay heat benchmarks for Spent Nuclear Fuel (one pincell and one assembly), perform calculations and compare and analyze the results in light of existing calorimetric measurements. The selected case is the PWR UO2 assembly 0E2, irradiated at the Ringhals-3 reactor and measured at the Clab facility in Sweden. In total, 21 institutes worldwide participated to the exercise, leading to 55 calculated results (named C). It was found that the measured decay heat values (E) can be satisfactorily reproduced with two-dimensional assembly calculations, leading to an average C/E value of 0.99, with an uncertainty (or one standard deviation) of ±0.01.
In this paper, we are estimating biases for the reactivity decrement and burnup of irradiated fuel assemblies, based on differences between measured and calculated local in-core power measurements (flux maps). The simulations are performed with validated CASMO5 and SIMULATE5 cycle follow-up models for three Pressurized Water Reactor (PWR) cores. A total of more than 70 cycles of operation and 1000 in-core flux maps at various core burnup steps are used, including both UO2 and MOX fuel assemblies, as well as UO2 alone, with relatively high assembly average burnup at the end of life (greater than 60 MWd/kg). The effect of nuclear data uncertainties is also calculated and quantified on both bias types. The main findings are that while the burnup biases are globally constant as a function of the assembly burnup, the bias for the reactivity decrement tends to increase, both in terms of average values and tolerance intervals.
Careful experimental uncertainty quantification (UQ) is key for developing trustworthy evaluated nuclear data. Templates to account for missing or under-reported experimental uncertainties were recently developed by the covariance committee of Cross Section Evaluation Working Group (CSEWG). In this work, we illustrate the practical application and limitations of these templates for selected neutron-induced reactions, including (n, tot), (n, γ), and (n, xn) in the fast energy range, to illustrate their use in data analyses for nuclear data evaluations. We show that while the templates provide consistent framework, proper implementation still requires detailed knowledge of experimental conditions and careful treatment of nonlinear effects in cross section derivation. Case studies highlight how template-assisted UQ improves consistency with previous evaluations such as ENDF/B and reveals open challenges in propagating uncertainties across different energy regimes. The main contribution of this paper is to connect formal template recommendations with their use in practical evaluation workflows, clarifying both their benefits and current limitations.
Small Modular Reactors (SMRs) represent a promising technology for addressing global energy demands. South Korea’s innovative SMR (i-SMR) is targeting soluble boron-free (SBF) operation to simplify reactor systems and eliminate boric acid corrosion. However, the elimination of soluble boron necessitates alternative reactivity control strategies to manage excess reactivity throughout extended fuel cycles, particularly for 30+ month operational periods that require higher initial fuel enrichment levels. This study evaluates the integration of three distinct burnable absorber (BA) approaches to achieve optimal reactivity control in LEU+ fuel configurations for extended cycle operation. The first approach employs Highly Intensive Gadolinia burnable Absorber (HIGA) rods containing concentrated Gd2O3 embedded in an alumina matrix, which achieves strong spatial self-shielding to extend reactivity control duration throughout the cycle. The second approach utilizes erbia (Er2O3) at natural isotopic composition uniformly admixed with fuel, providing smoother reactivity depletion characteristics through resonance-dominated burnup that supports long-cycle operation. The third approach involves enriched gadolinia mixed directly into fuel pins, leveraging high fractions of Gd-155 and Gd-157 isotopes to extend reactivity suppression while utilizing spatial self-shielding effects to slow absorber burnout. The synergistic combination of these three burnable absorber strategies enables effective management of the substantial excess reactivity inherent in LEU+ fuel while achieving cycle lengths exceeding 30 months. Core analysis is performed using the validated PRAGMA/SPHINCS nuclear design code system, where PRAGMA generates pin-level multigroup cross sections and depletion data, while SPHINCS performs three-dimensional pin-by-pin diffusion calculations. This approach to burnable absorber deployment maintains near-critical conditions throughout the extended cycle while avoiding the limitations of conventional single-absorber strategies that deplete too quickly in SBF operation. The successful demonstration of 35+ month fuel cycles through optimized burnable absorber integration significantly enhances SMR economic competitiveness by reducing refueling frequency, minimizing operational downtime, and improving overall plant capacity factors. These results provide a pathway for commercially viable SMR deployment with enhanced economic performance while maintaining the safety and operational flexibility advantages of soluble boron-free reactor designs.
Two industrial-scale methods for synthesizing nuclear-grade boron carbide (B4C) powder have been investigated and compared: direct synthesis from elements and carbothermal reduction. A series of experiments using different boron (isotopically enriched 10B and 11B) and carbon sources (carbon black and graphite powder) has been conducted. The synthesized powders have been analyzed for their chemical composition, purity, morphology, and particle size distribution. It was found that the carbothermal reduction, while producing chemically pure boron carbide, results in significant loss of boron due to the volatility of boron oxides, leading to a boron-deficient product with elongated, platelet-shaped particles. In contrast, direct synthesis from elements enables precise control over the B/C ratio and produces powders with polyhedral particles. However, the products of the direct synthesis method are often contaminated with impurities from the raw boron material, requiring subsequent mechanochemical treatment to achieve nuclear-grade purity. Both methods face challenges with morphological heterogeneity when scaled up, primarily due to temperature gradients within large powder volumes, which can also be addressed by mechanochemical treatment. As for loss of isotopic boron in the carbothermal process, it is 25–35%, while in direct synthesis it is 5–6%.
Small Modular Reactors (SMRs) and Advanced Modular Reactors (AMRs) are seen today as promising solutions for enhancing Europe’s energy supply security and contributing to global climate change mitigation. Their successful deployment, however, is contingent upon robust political support, widespread public acceptance, and the establishment of a harmonised safety regulatory framework. Furthermore, demonstrating their potential for non-electrical applications is crucial for stimulating broader interest and accelerating their development. Within the framework of the EURATOM Research and Development programme, these multifaceted challenges are concurrently addressed through the ECOSENS, HARMONISE, and SANE projects, whose results contribute to advancing knowledge pertaining to the societal, licensing, and safety challenges inherent in emerging nuclear energy technologies, with the aim to expedite their implementation within the European Union. The paper presents the up-to-date public perception on current and emerging nuclear technologies in the context of major societal challenges and details the outcomes of a comprehensive sustainability assessment of nuclear energy across its entire life cycle, employing a methodology co-developed by a diversity of stakeholders, including regulatory authorities, industry, academia, and civil society. Additionally, the paper introduces a novel economic model based on the System of Provision approach, designed to inform and assist decision-makers in formulating effective nuclear policies. These results provide a comprehensive perspective on nuclear energy, considering not only technical and economic aspects but also the broader societal and environmental implications. In addition, the paper provides recommendations for the harmonisation and standardisation of methodologies, codes and standards developed on the basis of the analysis of regulatory requirements for light water reactor technology. Finally, it outlines expectations arising from investigations into the safety aspects of non-electrical applications of nuclear energy, particularly focusing on residential heating, industrial processes, and other off-grid uses such as hydrogen production.
We analyze in-core reaction rate measurements coming from three Pressurized Water Reactors and two Boiling Water Reactors (named E) and compare them with CASMO5 and SIMULATE-3 (and -5) calculations (named C). A total of almost 1 million experimental values were considered, covering various cycles, detector systems and locations, as well as UO2 and MOX fuel types. Biases and root mean square values (rms) were extracted based on (E − C)/C ratios, leading to the finding of outliers, most likely due to detector issues and possible modeling effects for locations with low power. The average rms value is 2.9%, once the outliers are removed, but dissimilar values are obtained for different reactor types. The effect of nuclear data is also quantified for one PWR, leading to an absolute average rms uncertainty of 0.3%. We conclude, based on this study, that an improvement of the local rms values can be achieved by a better knowledge of measurements (i.e. uncertainties) and a possible update on specific parts of the involved models.
This work is conducted within the framework of nuclear fuel safety analyses involving multiscale modelling of UO2 fuel pellet fragmentation process. In this paper, a Finite Element (FE) methodology is proposed in order to enable 2D and 3D simulation of brittle fracture in micro-scale specimens used for the characterization of irradiated nuclear fuel. More specifically, these developments address Finite Element simulations using Cohesive Zone Modelling with complex shape of defects, where a direct application of Griffith’s criterion with the fracture mechanics approach is not always possible. In the case of UO2 ceramic material, the mesh refinement needed for the FE simulation of brittle fracture requires a specific attention due to the fact that the damage process zone has a characteristic size smaller than 10 nm. The main objectives are first to discuss the questions related to mesh convergence in the case of stress singularity, and secondly to enable the use of coarser meshes while maintaining accuracy compared to the converged mesh solution. The mesh convergence analysis and coarse mesh approach, established on a 2D configuration, are based on a Cohesive Zone Model (CZM) using two parameters: the critical cohesive stress σc, and the fracture energy Gc. The coarse mesh approach introduces a numerical critical cohesive stress, σc*, replacing the physical one, σc. Thanks to this approach, the relative mesh size can be increased by a ratio of ten with a deviation of the fracture load assessment smaller than 1% compared to the Griffith’s solution. The FE methodology and the coarse mesh approach are then tested with 2D and 3D simulations of a micro-cantilever bending test on a notched specimen. The simulation results are consistent with the experiment where the parameter Gc controls the fracture load. The critical stress, σc, ahead of the notch tip, can vary in the range 5–20 GPa with no effect on the fracture load. The coarse mesh approach is validated with a deviation lower than 2% compared to the fine converged mesh.
In the event of a severe accident in a light water nuclear reactor, substantial amounts of hydrogen may be produced and released into the containment due to reactor core degradation. Additionally, if molten corium interacts with concrete, further combustible gases, such as hydrogen and carbon monoxide, may be emitted. As demonstrated during the Fukushima Daichi accident, the combustion of H2 and CO can generate high-pressure peaks, potentially compromising the integrity of the reactor containment. To mitigate the formation of combustible gas mixtures, many European countries have implemented safety measures, including the use of Passive Autocatalytic Recombiners (PARs) or igniters, which help consume H2 and CO upon release. However, studies indicates that even with these systems in place, completely preventing the formation of combustible mixtures remains challenging. This could lead to localized combustion, flame acceleration, and ultimately poses a risk to the integrity of containment structures and safety components. To safeguard containment integrity, severe accident management guidelines (SAMG) recommend strategies such as activating spray systems or emergency core cooling, which may inadvertently increase the risk of gas explosions. Therefore, an effective gas monitoring system is of strong interest for supporting decision-making and implementing SAMG measures. In response to this need, the COMOS system – a fiber-coupled nuclearized gas probe prototype utilizing optical Raman technology – was developed under the French MITHYGENE Project. This system aims to enhance the management of combustible gases during severe accidents in water-cooled reactors, including Small Modular Reactors (SMRs), offering an innovative approach to improving nuclear safety.
The highly turbulent flow circulating between the rods inside a Pressurised Water Reactor (PWR) core creates turbulence induced vibrations. To get a better understanding of the flow characteristics, the French Alternative and Atomic Energy Commission (CEA) had built and operated CALIFS 5x5-test section. This mock up composed of a 5-by-5 rod bundle allowed velocity measurements using laser imaging techniques downstream an analytical spacer grid with and without mixing vanes. The CALIFS experiments and other experiments, revealed turbulence anisotropy downstream of spacer grids. The current study aims to assess the impact of spacer grid geometry on the turbulence anisotropy in a rod bundle using fine scale CFD simulation. The study focuses on the characterisation of the anisotropic behaviour of the flow in a rod bundle downstream simplified mixing grids inspired from the CALIFS 5x5 design. Wall-Resolved Large Eddy Simulations (WRLES) are used at a reduced Reynolds number of 14 000. The computation domain is limited to four subchannels around a central rod. Two simulations are performed, using different spacer grids designs (with and without mixing vanes). The results confirm mixing grid design strongly influences the overall turbulence anisotropy. In the case without mixing vanes, the impact of the grid on turbulence anisotropy seems to vanish 10 Hydraulic diameter (Dh) downstream of the grid. In the case with mixing vanes, the anisotropy remains in the subchannels up to 15 Dh downstream of the grid. In both cases, the turbulence anisotropy in the rod vicinity is strongly impacted close to the grid.
Turbulent heat transfer is a fundamentally complex physical process that has challenged turbulence modellers for many years. A full differential second-moment closure model provides a solid foundation for the derivation of the simpler Algebraic Heat Flux Model (AHFM), which retains the primary production terms that represent the underlying physical mechanisms driving the turbulent heat flux. This allows for accurate modelling of natural and mixed convection flows, which becomes particularly evident in advanced nuclear reactor applications with low Prandtl number coolants. In the present work, an AHFM model was implemented in the open-source code OpenFOAM, and was first verified using a simple channel test case. In the next step, the implementation was validated against Direct Numerical Simulation (DNS) data of a Rayleigh–Bénard convection case. The latter was based on simple geometry with natural convection of a low-Prandtl-number fluid, being of particular importance for Generation IV reactors using liquid metal coolants. The comparison of the generated numerical results with DNS data demonstrated the improvements in the prediction of turbulent heat flux in low-Prandtl number fluids with the new model.
The robotic complex “StarGate” was created in response to a request to design and deliver a device for the Nuclear Power Plant (NPP) V1 in Jaslovské Bohunice for the fragmentation of 12 steam generators. To meet the NPP’s requirements, a transdisciplinary team was assembled, consisting of experts in robotics, cybernetics, machining technologies, materials research, and nuclear energy. The synergy of specialists from all these key fields led to a unique solution with parameters and features that are beyond the current state of the art. The main differentiating parameters compared to the state of the art are as follows: Higher production efficiency and work productivity, comparable even to flame cutting (cutting speed of 1.7 mm/s with a wall thickness of 140 mm). The complete fragmentation of a single steam generator (140 tons) took 45 days. A higher level of process automation and intelligent optimization. Minimal radiation exposure for workers. Minimal aerosol generation and reduced spread of contamination to the surrounding environment. Shorter fragmentation time, thereby reducing overall costs (The complete fragmentation of a single steam generator (140 tons) took 45 days). According to today’s data, there are almost 100 steam generators in the world awaiting fragmentation and further processing (IAEA).
A U + 5 wt.% Al powder (written U-5Al), manufactured by centrifugal atomization, was used in this study to assess the feasibility of densifying UAlx + Al powder mixtures by Spark Plasma Sintering (SPS) and obtaining a final composition with a minimum content of UAl2 that meets specifications of targets for 99Mo production. The microstructure of the U-5Al powder was studied first and mainly consists of a mixture of UAl2 primary dendrites and alpha-U phase. Three SPS conditions differing in terms of UAlx/Al volume ratio and dwell time at 800 K were tested. The sintered discs were characterized by X-ray diffraction, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy. A very quick transformation of the U-5Al particles into a mixture of aluminides (UAl2 + UAl3 + UAl4, referred to UAlx) was evidenced: only a few minutes were sufficient to transform the U-based particles into a mixture of aluminides containing about 80 wt.% of UAl3. According to these first tests, this new highly versatile manufacturing process could advantageously replace the conventional method of producing targets, subject to technical adjustments and parametric studies.
We performed a high-precision study of the 215Po α-decay using a LaBr3 scintillating detector in a low-background environment. The 227Ac intrinsic contamination in the LaBr3 crystal undergoes a decay chain, producing the intermediate pair of 219Rn→215Po→211Pb decays. The fast time response and good energy resolution of the detector allow for extracting the short half-life of 215Po from the time correlation of the two subsequent α-decays by using the delayed coincidence method. Thanks to high statistics and a comprehensive uncertainty assessment, we obtain the most precise half-life value to date of 215Po, corresponding to 1.77804 ± 0.00091(stat.) ± 0.00023(syst.) ms. This measurement served the further purpose of benchmarking the half-life of its mother nucleus 219Rn, which is 3.921 ± 0.031(stat.) ± 0.083(syst.) s.
Many water-cooled SMRs (WC-SMRs) are being developed and their deployment is being considered in different countries world-wide, with some designs considering no soluble boron in the coolant. The elimination of soluble boron in the coolant of WC-SMRs offers advantages such as an enhanced negative moderator temperature coefficient, reduced corrosion, and simplified chemical systems, etc. leading to the growing interest in Soluble-Boron-Free (SBF) SMR-designs. Without soluble boron, excess reactivity control relies on burnable absorbers and control rods, potentially increasing power peaking factors and requiring careful core design to ensure safety margins. This study starts with an academic SBF Karlsruhe Small Modular Reactor (KSMR) core designed with a fresh fuel loading in order to find an optimal equilibrium cycle following a two-batch refueling strategy by optimizing fuel enrichment, burnable absorber rod configuration, and Gd2O3 content using CASMO5 and SIMULATE5. A Golang-based tool, named CoreOptimizer, was developed to automate the input file generation, simulation execution and output data extraction, enabling efficient core optimization. The resulting core design meets safety and performance criteria. This paper describes the optimization process, the applied tools and discusses the key neutronic and safety characteristics of the optimized equilibrium core design.
A U + 5 wt.% Al powder (written U-5Al), manufactured by centrifugal atomization, was used in this study to assess the feasibility of densifying UAlx + Al powder mixtures by Spark Plasma Sintering (SPS) and obtaining a final composition with a minimum content of UAl2 that meets specifications of targets for 99Mo production. The microstructure of the U-5Al powder was studied first and mainly consists of a mixture of UAl2 primary dendrites and α-U phase. Three SPS conditions differing in terms of UAlx/Al volume ratio and dwell time at 800 K were tested. The sintered discs were characterized by X-ray diffraction, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy. A very quick transformation of the U-5Al particles into a mixture of aluminides (UAl2 + UAl3 + UAl4, referred to UAlx) was evidenced: only a few minutes were sufficient to transform the U-based particles into a mixture of aluminides containing about 80 wt.% of UAl3. According to these first tests, this new highly versatile manufacturing process could advantageously replace the conventional method of producing targets, subject to technical adjustments and parametric studies.
Photoneutrons–neutrons produced through photonuclear reactions–play a critical role in radiation transport scenarios involving high-energy gamma sources, electron accelerators, and nuclear reactors. Accurate modeling of photoneutron production and transport is still a challenging problem, due to limitations in photonuclear data and the sensitivity of simulations to the choice of physics models. In this study, we present a comprehensive investigation of photoneutron generation using three state-of-the-art Monte Carlo codes–MCNP6®, PHITS, and TRIPOLI-4®–each coupled with two major nuclear data libraries: ENDF/B-VIII.1 and JENDL-5. Photon-induced neutron production is analyzed across 49 elemental targets for incident photon energies from the photonuclear threshold up to 30 MeV, encompassing the Giant Dipole Resonance (GDR) region. Key observables–photoneutron currents, energy spectra, and angular distributions–are systematically evaluated as a function of atomic number. The resulting dataset serves as a reference benchmark for simulation validation, informs future improvements in photonuclear modeling and nuclear data, and supports broader applications in theoretical and experimental nuclear physics.
