Journal of Nuclear Engineering

A Decay Heat Removal System (DHRS) is an essential passive safety feature in pool-type Sodium-Cooled Fast Reactors (SFRs), maintaining core temperatures within design limits via natural circulation after reactor scram. Operation of the DHRS is regulated by the damper of the Air Heat Exchanger (AHX), which controls its activation and shutdown. In the current design guidelines, it is typically recommended to initiate the Decay Heat Exchanger (DHX) at 600 s after a Station Blackout (SBO) event. However, this activation timing requires minor dynamic adjustment based on the transient response of the system, which can be obtained by either real-reactor experiments or numerical simulations. Since full-scale real-reactor experiments are not easy to conduct, numerical simulations are effective ways to enhance the passive safety performance of pool-type SFRs under SBO conditions, clarify the regulatory mechanism of DHX activation timing on system behavior, and optimize DHRS operational strategies. This study developed an integrated full-reactor three-dimensional numerical model that comprehensively incorporated key components such as the core, sodium pools, and DHX. Transient variations in power and boundary conditions were precisely controlled via User-Defined Functions (UDFs). The impact of different DHX activation strategies on the reactor’s decay heat removal capability was systematically analyzed. Three-dimensional numerical simulations were performed for three representative DHX operational strategies, immediate activation post-accident (0 s), delayed activation per the standard strategy (600 s), and complete DHX non-activation, yielding detailed temperature and flow field distributions within the reactor. Results demonstrate that under the standard strategy, not only can the temperature in the pool be controlled below the safety limit (550 °C) in the early stage but the temperature can also drop in the subsequent stage while retaining a 600 s safe operation threshold. Notably, the results reveal that “sooner is not always better”. Immediate DHX activation accelerates internal circulation and drives hot fluid downwards, paradoxically heating the cold pool faster than delayed activation, thereby resulting in a higher core outlet temperature. This study contributes to enhancing the credibility of passive safety in SFRs and provides reliable data to support the development of optimized reactor operation protocols. Full article

Gray-Box (GB) models are being considered for risk monitoring of Small Modular Reactors (SMRs). Their effectiveness is linked to the proper selection of the model parameters. This paper proposes a systematic methodology for identifying the most influential parameters of a GB model for estimating safety-critical variables of an SMR during normal operation and accident scenarios. The GB integrates a reduced-order physics-based model (White-Box, WB) with a data-driven (Black-Box, BB) model that corrects the outputs of the WB using the condition-monitoring data collected by sensors positioned onto the SMR. The proposed method combines signal decomposition, specifically the Hilbert–Huang Transform (HHT), and global sensitivity analysis (SA), based on first-order Kucherenko indices, to quantify the contribution of non-stationary, correlated GB input parameters to the variability of the safety-critical output parameters of interest. The proposed approach is applied to the Small Modular Dual Fluid Reactor (SMDFR), and the obtained results demonstrate its effectiveness in identifying informative and physically interpretable features, reducing complexity and computational burden to enable real-time risk monitoring. Full article

This review provides a critical and forward-looking analysis of established PET positron-emitting radionuclides—¹¹C (carbon-11),¹³N(nitrogen-13), ¹⁵O(oxygen-15), ¹⁸F(fluorine-18), ⁶⁸Ga (gallium-68),⁸²Rb(rubidium-82)—alongside some less widely adopted positron emitters—⁴⁴Sc (scandium-44), ⁶⁴Cu (copper-64), ⁸⁶Y (yttrium-86), ⁸⁹Zr (zirconium-89), ¹²⁴I(iodine-124)—examining the scientific, technological and operational factors influencing their clinical translation and applicability. Particular emphasis is placed on the role of nuclear properties as a key factor in radionuclide selection and development. For each radionuclide, the relevant aspects, including nuclear decay characteristics, production routes and logistical modalities, are discussed in terms of their impact on PET diagnostic performance and sustainability. The review summarizes recent technological advances designed to mitigate supply chain limitations that affect established positron emitters and discusses critical challenges related to other promising PET radionuclides, such as production scalability and dosimetric implications. Finally, ongoing developments in hybrid imaging platforms and multiparametric PET systems are briefly addressed, illustrating how these innovations are redefining diagnostic accuracy and accelerating the evolution of PET toward increasingly personalized clinical strategies. Full article

The Leibstadt Nuclear Power Plant (KKL) recently completed a comprehensive full-scope Fire Probabilistic Safety Assessment (Fire PSA) to fulfill the updated Swiss regulatory requirements (ENSI-A05) and align with international standards. The study was conducted using the NUREG/CR-6850 framework, incorporating state-of-the-art methodologies across different areas of the study, advanced fire modeling tools (CFAST and FDS), and the latest plant-specific data. As part of detailed fire modeling, a bespoke methodology was developed, tailored to KKL’s plant-specific characteristics, to ensure a systematic and standardized approach to fire scenario analysis while maintaining quality, consistency, and traceability. The analysis focused on evaluating fire risks in critical plant areas, such as the drywell, containment, main control room, remote shutdown areas, and cable spreading room. For each scenario, the fire-generated conditions, such as the extent of fire propagation and the time to damage targets, were analyzed using plant-specific heat release rate (HRR) and calorific potential (CALPOT) values. The study also addressed aspects such as multi-compartment analysis, fire-induced cable impacts, and treatment of multiple spurious operations. This paper highlights the methodological enhancements achieved by integrating international best practices and KKL-specific adaptations into a unified fire modeling framework. The results provide critical insights into fire propagation dynamics, validate the effectiveness of safety features, and support risk-informed decision-making for enhanced fire safety and regulatory compliance. The outcomes of fire modeling were utilized to develop fire event trees and refine the consequences of fire scenarios, thereby enabling a more realistic estimation of fire risk in the KKL Fire PSA study. Overall, the KKL PSA aims to serve as a benchmark for future fire risk assessments in the nuclear industry. Full article

In two-phase systems with heat transfer, developing tools that allow the analysis of interphase phenomena is crucial. In molten salt nuclear reactors, the fuel salt and helium in the core form a two-phase liquid–gas system. Understanding the heat transfer behavior between phases allows us to assess the impact of temperature changes in each phase as well as the feedback of neutron processes in the reactor. This work proposes using an upscaled heat transfer model to analyze the two-phase system, highlighting the importance of solving boundary value problems to obtain the closure variables in a unit cell with symmetry and periodicity. The closure variables are crucial for determining the heat transfer coefficients that exhibit the MSR’s scaled behavior. The coefficients are validated against the literature, and the results of the numerical experiments show that the cross-heat transfer coefficients exhibit symmetric properties. Full article

The steam generator is a core component of nuclear power plants that facilitates heat exchange between the primary and secondary circuits, directly impacting the overall operation of the plant in terms of safety and reliability. During prolonged operation, the heat transfer tubes of the steam generator are subjected to erosion, corrosion, and cracking due to high-temperature, high-pressure fluid impact and vibration. Existing in-service inspection strategies for heat transfer tubes generally employ fixed intervals and coverage, failing to effectively differentiate the actual risk of tubes in various regions, leading to wasted inspection resources or safety hazards. This paper proposes a dynamic inspection and plugging management strategy based on flow-induced vibration (FIV) analysis, specifically utilizing the flow stability ratio (FSR). By calculating the FSR of heat transfer tubes, the strategy categorizes them into high-risk, medium-risk, and low-risk regions, and dynamically adjusts inspection frequency and coverage based on these risk levels. Theoretical analysis and validation with actual data demonstrate that this strategy can improve inspection efficiency and ensure the safety of the steam generator. Full article

Fuel assembly design in Boiling Water Reactors has evolved to achieve more efficient use of uranium by optimizing the moderator distribution within the fuel assembly and increasing the number of smaller-diameter fuel rods to prevent rod power peaking. This evolution has gone from a 6-by-6 fuel rod arrangement to a 10-by-10 arrangement for the three major BWR fuel-assembly vendors. The designs of the fuel assemblies feature different radial and axial fuel rod distributions and inner water channels, with varying shapes and sizes. The main objective of these designs is to have a more homogeneous power distribution with a higher average burnup. The present study assesses the performance of these fuel assemblies, and the results show the impact of symmetry within the fuel assembly on the average enrichment and power distribution. Full article

In this study, an uncertainty quantification (UQ) and sensitivity analysis (SA) workflow was developed for the input parameters of the CANDLE module, which is currently being tested and verified for calculating the downward relocation and solidification of molten core material. The workflow consists of three steps: (i) Morris screening to reduce the input set, (ii) Sobol variance decomposition on the screened subset to compute Sobol sensitivity indices, and (iii) uncertainty propagation using a 2 × 2 design that combines two sampling schemes (MC and LHS) with two dependence settings (independent and correlated inputs). The four cases considered were independent MC, correlated MC, independent LHS, and correlated LHS–Iman–Conover (LHS-IC). We considered 16 input parameters and three output figures of merit (FOMs) and compared the four cases in terms of propagated uncertainty and Shapley-based importance rankings, thereby distinguishing the effects of the sampling scheme, the imposed input dependence, and their interaction. The results show that the molten mass of the current material in the source node is the dominant factor governing the drained melt mass and the remaining melt mass in the receiving node, whereas the cold-wall surface temperature has a significant effect on the mass of molten material that solidifies in the receiving node. The mass of molten material that remains available in the receiving node is mainly governed by the coupled effects of the molten mass of the current material at the source node, the length of the receiving node, and the velocity limit. Under the non-uniform input-parameter distributions adopted in this study, LHS broadened the range of the outputs. After input correlations were introduced, the output distributions changed slightly. This study improves the understanding of input parameter sensitivities and uncertainty propagation in the CANDLE module. It also demonstrates the practical use of LHS-IC for module-level UQ/SA with correlated inputs, providing guidance for subsequent model improvements and parameter tuning. Full article

The breeding blanket is a key component of tokamaks, primarily responsible for extracting heat from fusion reactions and for tritium breeding, which is essential to ensure a fusion reactor’s fuel self-sufficiency. Recent technological advancements have led to the development of Dual-Cooled Lead–Lithium (DCLL) breeding blankets, which employ a liquid metal (specifically a Lead–Lithium eutectic alloy) as a heat transfer medium and tritium breeder, while helium gas is used to cool the structural components of the reactor. The interaction between the moving electrically conducting fluid and the strong magnetic field in the tokamak environment leads to magnetohydrodynamic (MHD) effects. The latter are characterized by the induction of eddy currents within the fluid and resulting Lorentz forces generated by their interaction with the magnetic field, which cause additional pressure losses and reduce heat transfer efficiency. This work investigates the pressure drop experienced by a Lead–Lithium flow within a rectangular section conduit under the action of an external, uniform magnetic field of different intensities. An analytical model was developed to estimate the total MHD-induced pressure losses along the channel for different values of the external magnetic field intensity and then benchmarked against relative computational fluid dynamics (CFD) simulations carried out using COMSOL Multiphysics. This comparison allowed the validation of the analytical predictions as well as a better understanding of the influence of the applied magnetic field intensity on the overall pressure drop. Therefore, the aim of the analytical model is to provide analytical tools for reasonably accurate estimations of MHD pressure losses suitable for future preliminary design purposes. Full article

At the end of its last experimental campaign, in December 2023, the Joint European Torus (JET) became available for testing a compact and lightweight Laser-Induced Breakdown Spectroscopy (LIBS) system to be mounted on its robotic arm. The purpose of the test was the in situ chemical characterization of its internal walls and plasma-facing components (PFCs). Among the areas measured, special attention was devoted to the PFCs of the divertor, as this area is most affected by the re-deposition of material eroded from the first wall and unburned nuclear fuel (deuterium and tritium). In this article, we present the results of the LIBS characterization of a PFC of the High Field Gap Closure (HFGC), highly subjected to these phenomena. The in-depth distribution of several ITER-relevant chemical species is discussed through in-depth and correlation analyses, and the interpretation of the results is explained in terms of erosion and re-deposition of materials from the first wall. The study allowed us to estimate the thickness of the ablated layers by each laser shot, which is on the order of a few tens of nanometers, and to outline a mapping of the thickness of the re-deposited material. Full article

Plasma-facing materials (PFMs) for future fusion reactors require advanced mechanical and thermal properties to withstand the extreme challenges of high heat flux, plasma exposure, and neutron irradiation. Tungsten is one of the most suitable materials for use as a PFM in the divertor region. However, considering the high thermal loading/thermal stress combining plasma exposure and neutron irradiation/embrittlement, one of the major concerns for tungsten in PFMs is its intrinsic brittleness. To avoid cracking and components failure, tungsten toughening has been widely investigated, including the development of tungsten fiber-reinforced tungsten composites (W_f/W) using an extrinsic toughening mechanism, which could provide damage resilience against neutron embrittlement. Recently, a type of aligned long-fiber W_f/W (L-W_f/W) based on a powder metallurgical fabrication process was developed, demonstrating advanced fracture toughness while retaining other application-relevant properties. For L-W_f/W, the relatively easy production process suggests the feasibility and basis of industrialization. This work reports on the initial progress in industrializing L-W_f/W, with a focus on adapting the lab sintering process to a sintering process with industrial partner (Dr. Fritsch Sondermaschinen GmbH) and optimizing the process parameters. To improve the sinterability of tungsten and achieve higher density, various tungsten powders were explored, including commercial W powders, bimodal mixtures of different particle sizes, and granulated W powders. At the dedicated yttria interface, the thickness of yttria coating on the fibers was also optimized to ensure effective separation between the fibers and the matrix. Series of samples were produced with different dimensions up to 100 mm × 100 mm × 4 mm. After optimization, samples with 93% density and desired pseudo-ductility were prepared. Similarly to production in the lab, a major challenge in this work involved balancing the densification of the tungsten matrix with controlling fiber recrystallization and mitigating damage to the yttria interface. Full article

Qualification is the evidence-based process through which confidence is established that a component will perform its intended function, in its intended environment, for its intended lifetime, with the required reliability. It is an owner-led activity that defines the type, quantity and quality of data required for codification and for the industrial deployment of components and their structural materials. This paper presents a structured qualification framework and applies it to a fusion machine breeder blanket structure as a representative component. It demonstrates that qualification, rather than material properties alone, dictates the use of fusion structural materials and the deployment of such materials under ASME BPV and AFCEN RCC codes. Current limitations in addressing irradiation synergy, liquid metal corrosion, and joint integrity expose gaps that these codes cannot yet prescribe. Two contrasting structural blanket material case studies: metallic-based ferritic-martensitic steel Eurofer97 and non-metallic-based silicon carbide fibre-reinforced composites (SiC_f/SiC) are used to illustrate the differing evidence requirements for each system type. Industrial scale-up considerations, including alloy specifications, manufacturing readiness, inspection reliability, and supply-chain maturity, are evaluated alongside the need for internationally harmonised datasets and design methodologies. Fusion programmes can use a phased qualification strategy in which early, time-limited operation under controlled conditions builds the evidence needed for codification and scale-up, with the required pre-operation qualification level depending on risk, component criticality and failure consequences, and with the pace of qualification ultimately setting how quickly industry can supply components for commercial fusion. Codification remains essential for commercial deployment because construction codes express codified material behaviour through allowable stresses and permitted fabrication routes, enabling designers to use advanced materials without disclosing proprietary data. In jurisdictions where ASME BPV compliance is mandatory, codification determines whether a material may enter pressure boundary service and must therefore form part of the fusion machine owner’s long-term strategy for deployment. Full article

Silicon carbide (SiC) and SiC fiber-reinforced SiC matrix composites (SiC/SiC) are receiving renewed attention for use in next-generation fusion reactors due to their ability to withstand extreme conditions, including high temperatures, neutron irradiation, and plasma interactions. General Atomics Electromagnetic Systems (GA-EMS) has demonstrated significant progress in scaling up the fabrication of SiC/SiC, achieving high mechanical uniformity and meeting dimensional requirements in components up to 12 feet in length. Key developments are discussed including scale-up of the chemical vapor infiltration (CVI) process from lab-scale to full sized parts, high-dose (100 dpa) irradiation testing, nuclear-grade ceramic joining technologies, and production-focused quality control with the collective aim to establish SiC/SiC as a reliable solution for structural and functional components in fusion systems. Beyond manufacturing, the paper addresses supply chain barriers, particularly the limited availability and high cost of nuclear-grade SiC fiber. GA-EMS is developing a novel SiC fiber production method based on a thermochemical cure step that is anticipated to reduce costs compared to traditional approaches. Additionally, advancements in engineered SiC materials, such as SiC foams and tungsten-graded SiC composites, are discussed as promising solutions for specific fusion reactor components. Full article

During the operation of nuclear power plants, nuclear fuel undergoes significant compositional changes. After several cycles of use, the fuel must be removed and stored. Currently, spent fuel is stored mainly in pools or casks, and it is necessary to demonstrate the subcriticality of these systems. Spent nuclear fuel has a complex composition, and because computational codes are typically validated using fresh-fuel experiments, subcriticality assessments are usually performed conservatively with fresh-fuel compositions. These approaches demonstrate subcriticality but are very conservative and can lead to storage system designs that are more expensive or have reduced capacity. This paper focuses on the validation of computational codes using nuclear power plant critical start-up tests (referred to as reactor criticals). These tests include spent fuel and are well documented, allowing them to serve as validation experiments. Codes validated using reactor criticals can be applied to systems containing spent fuel calculation if sufficient similarity is demonstrated. Similarity is evaluated using the SCALE TSUNAMI-IP module, which is widely used for this purpose. Based on a database containing dozens of reactor criticals and similarity analyses, we developed a methodology for demonstrating the subcriticality of spent-fuel storage systems. Full article

High-temperature gas-cooled reactors (HTGRs) employ spherical fuel elements containing thousands of tristructural-isotropic (TRISO) particles, each centered on a UO₂ fuel kernel. The internal gelation process is a key technology for preparing these UO₂ fuel kernels. However, its application is limited by the poor room-temperature stability of conventional broths and the inherent trade-off between broth stability and mechanical strength. In this work, a novel five-component broth system composed of ZrO(NO₃)₂, hexamethylenetetramine (HMTA), urea, acetylacetone (ACAC), and glucose was developed. The synergistic effects of ACAC and glucose on broth stability and gelation kinetics were systematically investigated. An optimal ACAC/glucose molar ratio of 1:1 and an ACAC/ZrO₂⁺ ratio of 1.5 yielded a zirconium broth stable for over 5 h at 25 °C. Yttrium-stabilized zirconia (YSZ) microspheres prepared under optimized conditions exhibited excellent sphericity (1.04 ± 0.01), high density (5.84 g/cm³), and a crushing strength of 8.0 kg sphere⁻¹. Importantly, this stabilization strategy was successfully extended to the uranium broth, increasing its room-temperature stability from minutes to 6 h. The results demonstrate that the synergistic stabilization strategy effectively decouples the trade-off between broth stability and mechanical strength during the internal gelation process, providing an energy-efficient, scalable route for the preparation of nuclear fuel microspheres. Full article

Using factual information on background knowledge, costs, personnel numbers, resources, and facilities from the Manhattan Project, we examine the feasibility of the development of nuclear weapons in Germany in World War II. We conclude that, while for various reasons, a uranium bomb would have been technically and economically out of reach in Germany, a few plutonium bombs might have been possible had a coordinated aggressive project been initiated no later than about mid-1940. However, the German scientists involved never established an understanding of the functioning of an atomic bomb as contained in the Frisch–Peierls memorandum and were never asked to provide such a basis on which a decision on an atomic bomb program could be based. This means that a German atomic bomb program did not fail as is often assumed; rather, it was never started. The German uranium project was never more than a scientific mission to study the possibilities offered by the newly discovered source of nuclear power. Full article

The paper is focused on broader considerations regarding the commissioning process of the 3rd Unit of nuclear power plant VVER-440 type in Mochovce (Slovakia). The new nuclear plant built in Europe is getting much more slowly than expected, declared or scheduled. Besides the nuclear power plant in Olkiluoto (Finland) and also Flamanville (France), the 3rd Mochovce Unit has finally been in full operation since 6 November 2024. Nevertheless, the more than 30 years of construction process, which was intermittently stopped and frozen, make this success story exceptional. Lessons learned from commissioning are every time specific for different countries but commissioning of nuclear power plant without presence of general designer, respecting all safety requirements and taking full responsibility for this process is unique. Still, in general, the actual Slovak experiences and knowledge could help optimise new buildings in Europe, including dreams about small modular reactor deployment or the building of other clean and sustainable use of advanced nuclear facilities in the future. Full article

Molten Salt Reactors (MSRs) offer significant advantages over conventional reactors but introduce unique modeling challenges due to their circulating liquid fuel and strong coupling among nuclear, chemical, and fluid transport processes. These challenges are amplified in depletion calculations, where MSR specific phenomena such as online refueling, off-gas removal, material redistribution, and other flow driven processes must be accurately represented. This work presents a novel multi-point depletion model that efficiently and accurately predicts isotopic evolution in MSRs by explicitly accounting for these characteristics. The mathematical formulation is derived from first principles and is computationally implemented in the open-source depletion code ONIX using neutronics solutions from open-source transport code OpenMC. The new model represents the entire primary loop by dividing it into interconnected depletion zones and tracks nuclide transport, irradiation, and removal mechanisms through a system of coupled ordinary differential equations. This approach enables parallel computation and improves performance over traditional sequential depletion methods. Validation of the developed model against Molten Salt Reactor Experiment data shows good agreement for salt-seeking isotopes and those without noble gas precursors, while discrepancies for other nuclides suggest underestimation of the corresponding removal rates. The depletion model was further applied to a reference Molten Salt Fast Reactor design to assess a new reprocessing scheme intended to expedite the achievement of equilibrium operation. Full article