J. Nucl. Eng., Volume 7, Issue 2 (June 2026) – 11 articles

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