Journal of Nuclear Engineering

The sealing reliability of hydrogen-cooled generator systems in nuclear power plants is directly related to unit safety and outage critical-path optimization. Conventional airtightness pressure-holding tests usually use the 24 h leakage result as the acceptance criterion, but this occupies a long maintenance window. Early pressure and temperature signals are affected by thermal equilibration, environmental disturbances, and gas–oil coupling, making direct early assessment difficult. Based on historical pressure-holding test data from multiple nuclear power plants, this study develops a short-duration auxiliary assessment method. Test records from different plants are converted into a unified equivalent leakage rate, and a standardized dataset is established. A multi-branch framework is then developed, including leakage-trend prediction, local fluctuation identification, and feature-space validation. A conservative review strategy is introduced to support safety-oriented field decision making. The validation results show that the first 12 h monitoring data can support assessment of the 24 h leakage state. No false negatives were observed within the limited validation set. Samples with inconsistent outputs, near-threshold predictions, or abnormal feature-space locations are recommended for extended pressure holding and further review. Full article

Soluble-boron-free designs for water-cooled small modular reactors offer advantages such as reduced corrosion and simplified systems. However, the absence of soluble boron necessitates higher total control rod worth for reactivity control and the shutdown margin, leading to excessive individual control rod worth, which can lead to severe power excursions during a rod ejection accident (REA), potentially threatening the fuel integrity and core-cooling capability. The analysis of a hypothetical REA for an equilibrium core design showed that the fuel rod cladding failed due to the high reactivity worth of the ejected control rod. To enlarge the safety margins of this design under accidental conditions, two strategies were adopted: implementing a hybrid control rod configuration to decrease the local reactivity worth within single fuel assembly and re-arranging the refuelling loading pattern to prevent fresh fuel clustering. Using an in-house CoreOptimizer tool, the CASMO5 and SIMULATE5 simulations were automatized to find out an optimized equilibrium core design. The results demonstrated that all safety parameters of the optimized equilibrium core designs are within regulatory limits during normal operation and under REA conditions. By reducing the individual control rod worth, power spikes are considerably mitigated, thereby ensuring fuel integrity during an REA. Full article

Neutron noise in pressurized water reactors (PWRs) is mainly induced by fluctuations in macroscopic neutron cross-sections, which can be triggered by various factors such as vibrations of reactor internals. Existing studies mostly focus on calculation methods and software development of neutron noise, as well as in-core inversion diagnosis of noise source. Given the considerable differences in core design between Chinese PWRs CPR1000 and HPR1000, analyzing their propagation characteristics of neutron noise is significant for in-core anomaly detection and diagnosis of specific reactor types. This paper establishes a high-precision calculation method of neutron noise based on the transient neutron diffusion equation and Fourier transform technique. By simulating noise sources from macroscopic cross-section fluctuations, time-dependent relative power of each fuel assembly is obtained, and the amplitude and phase distribution of power fluctuations is derived via Fourier transform for propagation characteristic analysis. Simulations are conducted with assembly vibration noise sources for first-cycle and equilibrium-cycle cores of the HPR1000 and CPR1000. Numerical results indicate that propagation characteristics of core neutron noise are mainly dominated by noise source location and core configuration, with minor influence from burnup. Full article

Neoclassical tearing modes (NTMs) are magnetohydrodynamic instabilities that generate magnetic islands in tokamak plasmas, degrading confinement and potentially limiting high-performance operation. Their stabilization typically requires precise alignment and appropriate injection of electron cyclotron (EC) power beams, making real-time control a challenging task. In this work, we present a proof-of-principle study aimed at investigating the potential role of neural networks in the control of plasma instabilities. The objective is not the design of a controller for a specific machine, but rather to study how a learning-based agent can autonomously discover effective stabilization strategies through reinforcement learning. A synthetic environment based on a tokamak scenario is used as a test bed for this investigation; the specific scenario plays no essential role in the methodological conclusions. The controller is trained using reinforcement learning techniques and operates solely on a representation of the magnetic island width, without relying on equilibrium reconstruction or explicit knowledge of the deposition location relative to the island. Two control tasks are considered: pure angular alignment and combined angular alignment with power control. The strategies that autonomously emerge are consistent with hand-designed approaches reported in the literature, while the framework remains flexible for incorporating additional objectives such as power minimization. This exploratory study establishes a framework for assessing the potential advantages of data-driven approaches in magnetic island control and provides a basis for future investigations aimed at improving alignment and suppression strategies in fusion plasmas. Full article

The quantitative measurement of nuclear material hold-up in the process equipment is one of the technical challenges in nuclear material measurement for nuclear facilities. Its results are directly related to the optimization of radiation protection, the criticality safety control of nuclear materials, and the accurate accounting of nuclear material. As a key core equipment in the nuclear material reprocessing process, the precipitation reactor is restricted by the complex on-site environment, compact spatial layout, and continuous operation process, making it difficult for traditional measurement technologies to conduct accurate quantitative analysis of the internal hold-up. To address this issue, this paper proposes a method for measuring and analyzing the hold-up in the precipitation reactor based on the passive neutron counting method. A laboratory model of the precipitation reactor is constructed, and a multi-detector neutron measurement system is developed in this work. By combining Monte Carlo (MC) simulation with experimental calibration of standard point sources, a mathematical model suitable for hold-up measurement of the precipitation reactor is established. Meanwhile, uncertainty analysis of key data was carried out, and the accuracy of the model was verified by operational Pu samples of various masses, effectively reducing the measurement deviation caused by the uneven distribution of hold-up in the equipment and model assumptions. This research provides a more reliable technical reserve and reference paradigm for the measurement of nuclear material hold-up in nuclear facilities. Full article

Alkaline and alkaline earth metal chloride salts are used in molten chloride salt fast reactors (MCFRs) and pyro–electrometallurgical (or –electrochemical) recovering of uranium and transuranic elements (PERUT) from spent nuclear fuel. Reprocessing of MCFR spent fuel with the PERUT process, after recovery of U and transuranic elements (Np, Pu, Am, Cm), results in a chloride salt solvent waste stream containing fission and activation product chlorides. Recycling the chloride salt solvent by separation of fission and light element activation products (FPs and LEAPs) is highly desired because of the low chloride loading in the available glass and ceramic waste forms. This paper reviews the status of chloride salt waste management, chloride salt recycling studies, and potential FP and LEAP chlorides sequestration approaches. The chloride salt solvent recycling studies are represented by chemical precipitation of rare earth (RE) fission product chlorides with carbonate, O₂ gas and phosphate in LiCl and eutectic LiCl-KCl salt solvent, which is then followed by separation of Cs and Sr with distillation or crystallization. More than 99% removal efficiencies are attained for RE FP chlorides, and distillation removes more than 99% of Sr and Ba from the salt solvent. Volatile species released from the operation of MCFRs need to be sequestered. Minor chlorides species, such as SnCl₃, FeCl₃, CrCl₃, and ZrCl_2, will be present in the waste stream, and the separation of these species will be required for salt solvent recycling. Bromine and iodine can form bromides and iodides with metal elements such as alkaline and alkaline earth metal elements, which behave chemically similarly to their chloride counterparts. The presence of these compounds in the salt solvent waste may complexify the recycling process, for which more experimental studies are required. Full article

In molten salt reactors (MSRs), molten salt performs dual essential roles as fuel and coolant. The continuous circulation of the fuel salt in the primary loop inevitably leads to significant neutron activation of loop components, particularly the structural alloys of the heat exchanger (HX) and the coolant salt within the HX. This activation is strongly influenced by delayed neutron fluxes generated by the migration of delayed neutron precursors (DNPs) within the flowing fuel salt. Accurate quantification of the radioactivity of primary HX components is essential for establishing reliable modular replacement strategies, optimizing shutdown maintenance schedules, and ensuring operational safety. To address this requirement, a comprehensive simulation methodology has been developed to model the DNP transport through the primary HX in a small modular molten salt reactor (SM-MSR). It aims to quantitatively evaluate activation levels of HX structural alloys and circulating coolant salt within the HX. Comparative simulations were conducted to contrast scenarios with dynamic DNP migration and static-fuel scenarios excluding it. The results indicate that consideration of DNP migration increases the neutron flux at the top region of the HX by approximately three orders of magnitude compared with the static-fuel scenario. This elevates coolant salt radioactivity by over 50%. Significant increases in irradiation damage parameters (displacements per atom and helium production) are observed in the upper sections of HX structural alloys. These findings highlight the necessity of incorporating DNP migration effects for accurate prediction of primary loop component neutron activation. This provides a reference for future shielding design optimization, irradiation damage assessments, and shutdown dose rate calculations in the SM-MSR. Full article

During the life of a nuclear reactor, there are changes to the in-core components that are a function of operating environment and time. It is important to know how the properties of critical core components change, which can be assessed through materials surveillance programs. It is also desirable to characterize materials behaviour long before the end of the reactor design life. Therefore, experiments to characterize materials for in-core applications are performed in test reactors that typically have higher total neutron fluxes than power reactors. The extensive in-core materials irradiation programs that supported the validation of long-term material behaviour in CANDU (CANada Deuterium Uranium) reactors used various irradiation facilities, both domestic and international, are summarized in this paper. However, these test reactor facilities are aging and in some cases are closing, including NRU, which ceased operations in 2018. As Canada contemplates a new domestic high-flux test reactor to support both existing and potential new power reactors, this paper provides a review of the facilities and approaches that were implemented to successfully research CANDU reactor materials and can serve as a basis to define future facility requirements. Full article

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