Journal of Nuclear Engineering doi: 10.3390/jne7020035

Authors: Yue Liu Yuhao Zhang Ruoyu Liu Xinyi Chen Haijie Song Daogang Lu

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020034

Authors: Leonardo Miqueles Ibrahim Ahmed Francesco Di Maio Enrico Zio

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020033

Authors: Maria Letizia Terranova

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020032

Authors: Albena Tzenova Stoyanova Pavol Zvoncek Olivier Nusbaumer Devi Kompella Karthik Ravichandran Vignesh Anandan

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020031

Authors: Jesús Jorge Domínguez-Alfaro Alejandría D. Pérez-Valseca Gilberto Espinosa-Paredes Gustavo Alonso

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020030

Authors: Yi Yu Yicheng Zhang Lichen Tang Aimin Wu Chao Pian Yanfeng Qin Hao Wang Lushan Zhang

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020029

Authors: Hector Hernandez-Lopez Gustavo Alonso

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020028

Authors: Milica Ilic Piyush Sabharwall

It is our great pleasure to present this Special Issue on Advances in Thermal Hydraulics of Nuclear Power Plants [...]

Journal of Nuclear Engineering doi: 10.3390/jne7020027

Authors: Fenghui Yang Wanhong Wang Rubing Ma Xiaoming Yang

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020026

Authors: Silvia Iannoni Gianluca Camera Marcello Iasiello Nicola Bianco Giuseppe Di Gironimo

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020025

Authors: Salvatore Almaviva Lidia Baiamonte Jari Likonen Antti Hakola Juuso Karhunen Nick Jones Anna Widdowson Ionut Jepu Gennady Sergienko Rongxing Yi Rahul Rayaprolu Timo Dittmar Marc Sackers Erik Wüst Pavel Veis Shweta Soni Sahithya Atikukke Indrek Jõgi Peeter Paris Jasper Ristkok Pawel Gasior Wojciech Gromelski Jelena Butikova Sebastijan Brezinsek UKAEA RACE Team UKAEA RACE Team

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.

Journal of Nuclear Engineering doi: 10.3390/jne7020024

Authors: Yiran Mao Ute Wilkinson Jan Willem Coenen Daniel Wilkinson Johann Riesch Christian Linsmeier

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 (Wf/W) using an extrinsic toughening mechanism, which could provide damage resilience against neutron embrittlement. Recently, a type of aligned long-fiber Wf/W (L-Wf/W) based on a powder metallurgical fabrication process was developed, demonstrating advanced fracture toughness while retaining other application-relevant properties. For L-Wf/W, the relatively easy production process suggests the feasibility and basis of industrialization. This work reports on the initial progress in industrializing L-Wf/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.

Journal of Nuclear Engineering doi: 10.3390/jne7010023

Authors: Emily R. Lewis Guy Anderson Diego Martinez de Luca Bradley A. Young Thomas P. Davis

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 (SiCf/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.

Journal of Nuclear Engineering doi: 10.3390/jne7010022

Authors: George M. Jacobsen Sean Gonderman Rolf Haefelfinger Lucas Borowski Ivan Ivanov William McMahon Jiping Zhang Osman Trieu Christian P. Deck Hesham Khalifa Tyler Abrams Zachary Bergstrom Christina A. Back

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010021

Authors: Matej Sikl Radim Vocka

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010020

Authors: Rui Xu Xiao Yuan Jianjun Li Changsheng Deng Ziqaing Li Xingyu Zhao Shaochang Hao Bing Liu Yaping Tang Jingtao Ma

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010019

Authors: Manfred Popp Piet de Klerk Bruce Cameron Reed

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010018

Authors: Vladimír Slugeň Gabriel Farkas Jana Šimeg Veterníková Slavomír Bebjak Peter Andraško Martin Mráz

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010017

Authors: Mohamed H. Elhareef Zeyun Wu

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.

Journal of Nuclear Engineering doi: 10.3390/jne7010016

Authors: Lucrezia Spagnuolo Luigi Lepore Simone Placidi Francesca Limosani Angela Pagano Tiziana Guarcini Francesca Varsano Luca Falconi Valentina Fabrizio Davide Formenton Andrea Roberti Marco Capogni

The global demand for medical radionuclides is rapidly increasing, driven by the expansion of diagnostic and therapeutic radiopharmaceuticals, and by recurrent vulnerabilities in international supply chains. While high-flux reactors remain the backbone of large-scale isotope production, low- and medium-power research reactors—such as TRIGA facilities—offer valuable opportunities for decentralised, flexible, and alternative radionuclide generation. Several studies have demonstrated their capability to produce emerging therapeutic or diagnostic isotopes, including 111Ag, 99Mo/99mTc, 64Cu, 177Lu, and 161Tb, although with yield limitations inherent to moderate neutron flux levels. In Europe, recent initiatives such as PRISMAP, SIMPLERAD, and SECURE aim to strengthen production capacity and diversify radionuclide availability. Within this framework, Italy—lacking operational power reactors—seeks alternative routes to ensure a resilient national supply. This work presents the investigation carried out within the SECURE project to assess the feasibility of an Italian production cycle for medical-grade 161Tb at the ENEA TRIGA RC-1 Research Reactor (Rome). The study integrates reactor-specific irradiation analyses, the development of chemical separation and target recovery processes, and a comprehensive economic evaluation within a full lifecycle perspective. The results highlight the potential and constraints of a TRIGA-based production for supporting future Italian theranostic needs.

Journal of Nuclear Engineering doi: 10.3390/jne7010015

Authors: Akshay Kumar Khandelwal Charie A. Tsoukalas Yang Zhao Mamoru Ishii

Objective neural network-based two-phase flow regime classifiers are developed for vertical, narrow, rectangular channels and a 1 inch round pipe using Kohonen Self-Organizing Maps. In the rectangular channel, the classifier uses five geometric inputs obtained from a two-sensor droplet-capable conductivity probe (DCCP-2): the bulk gas void fraction αg, ligament void fraction αlig, normalized ligament chord length ylig, normalized large bubble chord length yℓ,bb, and a droplet indicator. These parameters allow for the objective identification of bubbly/distorted bubbly, cap-turbulent, churn-turbulent, annular, rolling wispy, and wispy flow regimes, and yield quantitative transition boundaries in the (jf,jg) plane for a densely populated test matrix. In the round pipe, a four-sensor droplet-capable conductivity probe (DCCP-4) provides the mean and standard deviation of droplet, bubble, and ligament chord length distributions, which are used as inputs to a Self-Organizing Map (SOM) classifier that separates rolling annular and wispy annular regimes at high void fractions. The resulting regime maps are discussed in terms of the associated phase geometries and their impact on interfacial area, drag, and entrainment, providing regime-dependent geometric inputs that can be used to improve Two-Fluid Model closures for reactor downcomers, core channels, and other nuclear thermal–hydraulic applications.

Journal of Nuclear Engineering doi: 10.3390/jne7010014

Authors: Scott D. Ramsey

The neutron survival probability (and related quantities including probabilities of extinction and initiation) is a central element of the broader stochastic theory of neutron populations and finds application in fields including reactor start-up, analysis of reactor power bursts and criticality accidents, and safeguards. In a full neutron transport formulation, the equation governing the single-neutron survival probability is a backward or adjoint-like integro-partial differential equation with the added complexity of being highly nonlinear. Analogous formulations of this equation exist in the context of many approximate theories of neutron transport, with the point kinetics formulation having received significant theoretical attention since the 1940s. This work continues this tradition by providing a novel analysis of the single-neutron survival probability equation using the tools of boundary layer theory. The analysis reveals that the “fully dynamic” solution of the single-neutron survival probability equation—and some key probability distributions derived from it—may be cast as a singular perturbation around the underlying quasi-static single-neutron probability of initiation. In this perturbation solution, the expansion parameter is the ratio of the neutron generation time to a macroscopic time scale characterizing the overall system evolution; this interpretation illuminates some of the fundamental structural aspects of neutron survival phenomena.

Journal of Nuclear Engineering doi: 10.3390/jne7010013

Authors: Kang Seog Kim Matthew Jessee Andrew Holcomb William Wieselquist

The SCALE transport lattice code, Polaris, has been previously developed to generate few-group homogenized cross sections for whole-core nodal diffusion simulators in which the embedded self-shielding method (ESSM) is used for resonance self-shielding calculations to process cross sections. Although the ESSM capability has been very successful in light-water reactor analysis, it may require enhancements in computational efficiency; treatment of spatially dependent resonance self-shielding effects; and handling of interrelated resonance effects among fuel, cladding, and control rod materials. Therefore, this study focuses on improving computational efficiency by using a Dancoff-based Wigner–Seitz approximation combined with a material-based resonance categorization, through which a spatially dependent ESSM capability is developed to accurately estimate self-shielded cross sections inside the fuel. Benchmark results show that the new capability significantly enhances computational efficiency and accuracy for spatially dependent local zones within the fuel and through depletion.

Journal of Nuclear Engineering doi: 10.3390/jne7010012

Authors: Julian Duran-Gonzalez Victor Hugo Sanchez-Espinoza

KANECS is a 3D multigroup neutronics code based on the Simplified Spherical Harmonics (SPN) approximation and the Continuous Galerkin Finite Element Method (CGFEM). In this work, the code is extended to solve the time-dependent neutron kinetics by implementing a fully implicit backward Euler scheme for the neutron transport equation and an implicit exponential integration for delayed neutron precursors. These schemes ensure unconditional stability and minimize temporal discretization errors, making the method suitable for fast transients. The new formulation transforms each time step into a transient fixed-source problem, which is solved efficiently using the GMRES solver with ILU preconditioning. The kinetics module is validated against established benchmark problems, including TWIGL, the C5G2 MOX benchmark, and both 2D and 3D mini-core rod-ejection transients. KANECS shows close agreement with the reference solutions from well-known neutron transport codes, with consistent accuracy in normalized power evolution, spatial power distributions, and steady-state eigenvalues. The results confirm that KANECS provides a reliable and accurate framework for solving neutron kinetics problems.

Journal of Nuclear Engineering doi: 10.3390/jne7010011

Authors: Yuqi Liu Minghui Chen

This study evaluates the performance and scalability of fluoride-salt-lubricated hydrodynamic journal bearings used in primary pumps for Fluoride-salt-cooled High-temperature Reactors (FHRs). Because full-scale pump prototypes have not been tested, a scaling analysis is used to relate laboratory results to commercial conditions. Bearings with different length-to-diameter (L/D) ratios were assessed over a range of shaft speeds to quantify geometric and hydrodynamic effects. High-temperature bushing test data in FLiBe at 650 °C were used as inputs to three-dimensional computational fluid dynamics (CFD) simulations in STAR-CCM+. Applied load, friction force, and power loss were computed across operating speeds. Applied load increases linearly with shaft speed due to hydrodynamic pressure buildup, while power loss increases approximately quadratically, indicating greater energy dissipation at higher speeds. The resulting correlations clarify scaling effects beyond small-scale testing and provide a basis for bearing design optimization, prototype development, and the deployment of FHR technology. This work benchmarks speed-scaling relations for fluoride-salt-lubricated hydrodynamic journal bearings within the investigated regime.

Journal of Nuclear Engineering doi: 10.3390/jne7010010

Authors: Yunlong Jia Long Guo Wei Li Shuai Zhang Xiaojie Shi Shengming Yin

The performance of conventional precipitation-strengthened copper alloys drastically degrades at temperatures exceeding 500 °C, hindering their application under extreme conditions like those in nuclear fusion reactors. Oxide dispersion–strengthened copper (ODS–Cu) alloy surmounts these constraints by incorporating thermally stable, nanoscale oxide dispersoids that simultaneously confer strengthening, microstructural stabilization, and enhanced irradiation tolerance, while preserving high thermal conductivity. This review comprehensively examines the state of the art in ODS–Cu alloy from a “processing–microstructure–property” perspective. We critically assess established and emerging fabrication routes, including internal oxidation, mechanical alloying, wet chemical synthesis, reactive spray deposition, and additive manufacturing, to evaluate their efficacy in achieving uniform dispersions of coherent/semi-coherent nano-oxides at engineering-relevant scales. The underlying strengthening mechanisms and performance trade-offs are quantitatively analyzed. The review also outlines strategies for joining and manufacturing complex components, highlights key gaps in metrology and reproducibility, and proposes a roadmap for research and standardization to accelerate industrial deployment in plasma-facing components.

Journal of Nuclear Engineering doi: 10.3390/jne7010009

Authors: C. Erika Moss Ondrej Chvala Donny Hartanto

Targeted radionuclide therapy (TRT) is an innovative and flexible approach for treating various forms of cancer, enabling selective delivery of cytotoxic radiation to cancerous cells while minimizing damage to healthy tissue. Although TRT has proven to be highly promising for treating even advanced-stage cancers, ensuring a stable supply of the radionuclides essential for its use remains a significant challenge today. This is also true for radionuclides utilized in nuclear imaging procedures, such as Positron Emission Tomography (PET) and Single Photon Emission Computed Tomography (SPECT). Liquid-fueled molten salt reactors (MSRs) are promising for producing large quantities of highly desirable radionuclides for imaging and therapy, offering the ability to recover these radionuclides online without the need for interruptions to power production. In this study, the production of numerous beta- and alpha-emitting radionuclides for use in TRT and diagnostic procedures was studied in two small, geometrically identical, thermal spectrum MSR models—one operating with LEU fuel, and the other with a mixture of HALEU and thorium—using a novel MSR refueling and waste management concept. For therapeutic alpha emitters such as 225Ac and 213Bi, the impact of thorium utilization on production yields was significant, facilitating greatly increased production.

Journal of Nuclear Engineering doi: 10.3390/jne7010008

Authors: Fergany Badry Merve Gencturk Karim Ahmed

Accurate prediction of the effective thermal conductivity (ETC) of nuclear fuels is essential for optimizing fuel performance and ensuring reactor safety. However, the experimental determination of ETC is often limited by cost and complexity, while high-fidelity simulations are computationally intensive. This study presents a novel hybrid framework that integrates experimental data, validated mesoscale finite element simulations, and machine-learning (ML) models to efficiently and accurately estimate ETC for advanced uranium-based nuclear fuels. The framework was demonstrated on three fuel systems: UO2-BeO composites, UO2-Mo composites, and U-10Zr metallic alloys. Mesoscale simulations incorporating microstructural features and interfacial thermal resistance were validated against experimental data, producing synthetic datasets for training and testing ML algorithms. Among the three regression methods evaluated, namely Bayesian Ridge, Random Forest, and Multi-Polynomial Regression, the latter showed the highest accuracy, with prediction errors below 10% across all fuel types. The selected multi-polynomial model was subsequently used to predict ETC over extended temperature and composition ranges, offering high computational efficiency and analytical convenience. The results closely matched those from the validated simulations, confirming the robustness of the model. This integrated approach not only reduces reliance on costly experiments and long simulation times but also provides an analytical form suitable for embedding in engineering-scale fuel performance codes. The framework represents a scalable and generalizable tool for thermal property prediction in nuclear materials.

Journal of Nuclear Engineering doi: 10.3390/jne7010007

Authors: Hiroyuki Yamada Masato Nakajima Hiromichi Miura Ryusuke Haraguchi Yoshinori Mihara Eishiro Higo

Based on lessons learned from the Fukushima Daiichi Nuclear Power Plant accident caused by the 2011 off the Pacific coast Tohoku Earthquake, and the subsequent tsunamis, Japanese utilities have been upgrading their tsunami countermeasures. To understand the residual risk from beyond-design-basis events, it is important to assess seismic and tsunami risks independently while also recognizing how a plant’s risk profile changes when these events occur concurrently. This study developed a framework for a multi-hazard probabilistic risk assessment (PRA) to evaluate risks to nuclear power plants (NPPs) resulting from the superposition of earthquake and tsunami events. The framework is proposed on the assumption that the targeted plant has previously conducted single-hazard PRAs for both earthquakes and tsunamis. This study presents an approach to define the scope of risk assessment for the superposition of earthquake and tsunami events, based on the results from single-hazard PRAs for each hazard. It provides an analytical framework for superposition scenario analysis and a simplified method for multi-hazard assessment using data from single-hazard assessments. Moreover, a series of steps for the multi-hazard fragility assessment of superposed earthquake and tsunami events are proposed, clarifying the relationship between superposed impacts and the damaged parts and damage modes, accompanied by illustrative examples.

Journal of Nuclear Engineering doi: 10.3390/jne7010005

Authors: Zhihong Tang Bin Yang Wenjun Zhang Ruohan Chen Shusheng Guo Junfeng Li Liyuan Wang Xing Huang

Lead–bismuth eutectic (LBE), while advantageous for advanced nuclear reactors due to its thermophysical properties, faces oxidation and corrosion challenges during operation. This study aims to optimize gas-phase oxygen control by computationally analyzing oxygen transport dynamics in an LBE loop. High-fidelity simulations were performed using ANSYS Fluent and STAR-CCM+ based on the CORRIDA loop geometry, employing detailed meshing for convergence. Steady-state analyses revealed localized oxygen enrichment near the gas–liquid interface (peaking at ∼3×10−6 wt%), decreasing to ∼5.0−6.8×10−8 wt% at the outlet. Transient simulations from an oxygen-deficient state (1×10−8 wt%) demonstrated distribution stabilization within 150 s, driven by convection-enhanced diffusion. Parametric studies identified a non-monotonic relationship between inlet velocity and oxygen uptake, with optimal performance at 0.7–0.9 m/s, while increasing temperature from 573 K to 823 K monotonically enhanced the outlet concentration by >200% due to improved diffusivity/solubility. The average mass transfer coefficient (0.6–0.7) aligned with literature values (±20% deviation), validating the model’s treatment of interface thermodynamics and turbulence. These findings the advance mechanistic understanding of oxygen transport in LBE and directly inform the design of oxygenation systems and corrosion mitigation strategies for liquid metal-cooled reactors.

Journal of Nuclear Engineering doi: 10.3390/jne7010006

Authors: Maria Letizia Terranova

Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic modality which couples selective pharmacological delivery of 10B with irradiation by low-energy neutrons to achieve highly localized tumor cell killing. The BNCT therapeutic approach is undergoing rapid evolution driven primarily by advances in compact accelerator-driven neutron-source and associated facility-level nuclear infrastructure. This review examines the key physical and radiobiological principles of BNCT, with emphasis on the current engineering and operational aspects, such as neutron production and moderation, spectral shaping, beam optimization and dosimetric quantification, that critically influence clinical translation. Recent progress in 10B production and enrichment, as well as in strategies for efficient 10B delivery, is also briefly addressed. By tracing the pathway from neutron source to clinical target, this review defines the state of the art in BNCT technology, identifies the main physical and infrastructural challenges, and delineates the multidisciplinary advances needed to support widespread clinical implementation of next-generation BNCT systems.

Journal of Nuclear Engineering doi: 10.3390/jne7010004

Authors: Shasha Lv Yingtao Ma Chong Tian Jie Gao Yumeng Zhao Zhengcao Li

The structural integrity of nuclear graphite is paramount for the lifespan of High-Temperature Gas-Cooled Reactors. The nuclear graphite components operate under extreme conditions involving high temperature, pressure, and intense neutron irradiation, leading to complex service behavior that is difficult to characterize only by experimental methods. This study employs the finite element method (FEM) to assess component stress and failure risk. The ManUMAT simulation method was first validated against irradiation data for Gilsocarbon graphite from an Advanced Gas-Cooled Reactor and was subsequently applied to stress–strain analysis of the nuclear graphite bricks in the HTR-PM side reflector layer. The 3D micropore structure of nuclear graphite was obtained via X-μCT and reconstructed in Avizo to establish an FEM model based on the actual pore geometry. Simulations of nuclear graphite over a 30 full-power-year service period predicted a significant contraction on the core-side and minimal thermal expansion on the out-side driven by the neutron doses. This research establishes a finite element framework that extends the ManUMAT approach by integrating a realistic pore structure model, thereby providing a foundation for quantifying the microstructural effects on macroscopic performance.

Journal of Nuclear Engineering doi: 10.3390/jne7010003

Authors: Anas Muhamad Pauzi Wenyu Mao Andrea Cioncolini Eddie Blanco-Davis Hector Iacovides

Fretting wear due to flow-induced vibration (FIV) remains a primary cause of fuel failure in light water nuclear reactors. In the study of axial FIV, i.e., FIV caused by axial flows, three vibration characteristics, namely natural frequency, damping ratio, and root-mean-square (RMS) amplitude, are critical for mitigating fretting wear by avoiding resonance, maximising overdamping, and preventing large-amplitude instability motion, respectively. This paper presents a set of best practices for simulating axial FIV with a focus on predicting these parameters based on a URANS-FSI numerical framework, utilising high-Reynolds-number Unsteady Reynolds-Averaged Navier–Stokes (URANS) turbulence modelling and two-way fluid–structure interaction (FSI) coupling. This strategy enables accurate and efficient prediction of vibration parameters and offers promising scalability for full-scale nuclear fuel assembly applications. Validation is performed against a semi-empirical model to predict RMS amplitude and experimental benchmarking. The validation experiments involve two setups: vibration of a square beam with fixed and roller-supported ends in annular flow tested at Vattenfall AB, and self-excited vibration of a cantilever beam in annular flow tested at the University of Manchester. The study recommends best practices for numerical schemes, mesh strategies, and convergence criteria, tailored to improve the accuracy and efficiency for each validated parameter.

Journal of Nuclear Engineering doi: 10.3390/jne7010002

Authors: Ashhar Bilal Puzhen Gao Muhammad Irfan Khalid Abid Hussain Ali Mansoor

The temperature fluctuations due to the mixing of two streams in a T-junction induce thermal stresses in the piping material, resulting in a pipe failure in Nuclear Power Plants. The numerical modeling of the thermal mixing in T-junctions is a challenging task in computational fluid dynamics (CFD) as it requires advanced turbulence modeling with scale-resolving capabilities for accurate prediction of the temperature fluctuations near the wall. One approach to address this challenge is using Partially Averaged Navier–Stokes modeling (PANS), which can capture the unresolved turbulent scales more accurately than traditional Reynolds-Averaged Navier–Stokes models. PANS modeling with k-ε closure gives encouraging results in the case of the Vattenfall T-junction benchmark case. In this study, PANS k-ω closure modeling is implemented for the WATLON T-junction Benchmark case. The momentum ratio (MR) for two inlet streams is 8.14, which is a wall jet case. The time-averaged and root mean square velocity and temperature profiles are compared with the PANS k-ε and LES results and with experimental data. The velocity and temperature field results for PANS k-ω are close to the experimental data as compared to the PANS k-ε for a given filter control parameter fk.

Journal of Nuclear Engineering doi: 10.3390/jne7010001

Authors: David Bowden Benjamin Evans Jack Haley Jim Johnson Alexander Carruthers Stephen Jones Dane Hardwicke Talal Abdullah Shahin Mehraban Nicholas Lavery Paul Sukpe Richard Birley Abdollah Bahador Alan Scholes Peter Barnard

Future fusion power plants require structural materials that can withstand extreme operating conditions, including high coolant outlet temperatures, mechanical loading, and radiation damage. Reduced-activation ferritic martensitic (RAFM) steels are a primary candidate as a structural material for such applications. This study demonstrates the successful production of a 5.5-tonne RAFM billet via electric arc furnace (EAF) technology, enabling scalable, cost-effective manufacturing. The resulting UK-RAFM alloy offers superior tensile strength and creep lifetime performance compared to Eurofer97. This is attributed to alterations in the initial forging process during manufacture. Modified thermomechanical treatments (TMTs) were subsequently applied to the UK-RAFM, which are shown to enhance the tensile strength further, particularly at 650 °C. Building on this, an Advanced RAFM (ARAFM) steel was designed to exploit the benefits of optimised chemistry to encourage metal carbonitride (MX) precipitate evolution alongside bespoke TMTs. Challenges around ensuring suitable processing windows in these steels, to avoid the over-coarsening of MX precipitates or the formation of deleterious delta-ferrite, are discussed. A subsequent 5.5-tonne ARAFM billet has since been produced using EAF facilities, with performance to be reported separately. This work highlights the synergy between alloy design, process optimisation, and industrial scalability, paving the way for a new generation of low-cost, high-volume, fusion-grade steels.

Journal of Nuclear Engineering doi: 10.3390/jne6040056

Authors: Gianmarco Grippo Calogera Lombardo Massimiliano Polidori

In recent decades, international interest has grown in the design and implementation of evolutionary reactors based on passive systems. The design of such systems requires reliable and validated numerical tools capable of simulating phenomena driven by very small forces, especially when compared to active systems. For this reason, several international research projects aim to assess the capabilities and limitations of numerical tools in modelling passive systems and their associated physical phenomena. The HERO-2 facility was designed to provide preliminary experimental data for characterizing bayonet tubes and exploring their potential application as Steam Generators (SGs) in advanced nuclear reactor designs, such as Small Modular Reactors (SMRs). Following the agreement between the Italian Ministry of Economic Development and the ENEA, multiple experimental campaigns were conducted, and a RELAP5 (R5) input deck of the facility has been developed. Considering the RELAP5 limits in simulating condensation phenomena encountered in previous studies, the primary objective of this study is to enhance the capabilities of the code in simulating condensation phenomena in horizontal pipes under natural circulation conditions with the implementation of Thome correlation and, in the second instance, to re-evaluate the numerical model of the HERO-2 facility. Moreover, a comprehensive uncertainty analysis (UA) is carried out to identify the key parameters influencing the simulations. The analysis revealed that the simulation results are strongly affected by the filling ratio uncertainties, a given initial condition that, together with the power supplied, determines the most important thermal-hydraulic (T/H) test parameters, such as the saturation pressure, the void fraction, mass flow rate, etc. Overall, the study provides a deeper understanding of the factors governing passive system performance and highlights the importance of accurately characterizing the experimental boundary and initial conditions in the verification and validation activities of a T/H code.

Journal of Nuclear Engineering doi: 10.3390/jne6040055

Authors: Ivan E. Novoselov Ivan S. Zhidkov

Lead halide perovskites, exemplified by methylammonium (MA) lead iodide (MAPbI3), combine strong optical absorption, long carrier diffusion lengths, and defect-tolerant electronic structure with facile processing, making them attractive for photovoltaics and radiation detection. Yet, their behavior under electron irradiation remains insufficiently understood, limiting deployment in space and dosimetry contexts. Here, we employ Monte Carlo simulations (Geant4) to model electron interactions with MAPbI3 across energies from 0.1 to 100 MeV and absorber thicknesses from 10 μm to 1 cm. We quantify deposited energy, event statistics, energy per interaction, non-ionizing energy loss, and dominant radiation effects. The results reveal strong thickness-dependent regimes: thin photovoltaic-type layers (~hundreds of nanometers) are largely transparent to MeV electrons, minimizing bulk damage but allowing localized ionization, exciton self-trapping, and photoexcitation-driven ion migration. Although localized excitations can temporarily improve carrier collection under short-term exposure, their cumulative effect drives ionic rearrangement and defect growth, ultimately reducing device stability. In contrast, thicker detector-type films (10–100 μm) sustain multiple scattering and ionization cascades, enhancing sensitivity but accelerating defect accumulation. At centimeter scales, energy deposition saturates, enabling bulk-like absorption for high-flux dosimetry. Overall, electron irradiation in MAPbI3 is dominated by electronic excitation rather than ballistic displacements, underscoring the need to optimize thickness and composition to balance efficiency, sensitivity, and durability.

Journal of Nuclear Engineering doi: 10.3390/jne6040054

Authors: Mohamed S. El-Genk Timothy M. Schriener Ahmad N. Shaheen

This work investigated machine-learning algorithms for remote-control and autonomous operation of the Very-Small, Long-Life, Modular (VSLLIM) microreactor. This walk-away safe reactor can continuously generate 1.0–10 MW of thermal power for 92 and 5.6 full power years, respectively, is cooled by natural circulation of in-vessel liquid sodium, does not require on-site storage of either fresh or spent nuclear fuel, and offers redundant means of control and passive decay heat removal. The two ML algorithms investigated are Supervised Learning with Long Short-Term Memory networks (SL-LSTM) and Soft-Actor Critic with Feedforward Neural Networks (SAC-FNN). They are trained to manage the movement of the control rods in the reactor core during various transients including startup, shutdown, and to change the reactor steady state power up to 10 MW. The trained algorithms are incorporated into a Programmable Logic Controller (PLC) coupled to a digital twin dynamic model of the VSLLIM microreactor. Although the SL-LSTM algorithms demonstrate high prediction accuracy of up to 99.95%, they demonstrate inferior performance when incorporated into the PLC. Conversely, the PLC with SAC-FNN algorithm accurately adjusts the control rods positions during the reactor startup transients to within ±1.6% of target values.

Journal of Nuclear Engineering doi: 10.3390/jne6040053

Authors: Bayan Kurbanova Yuriy Sizyuk Ansar Aryngazin Zhanna Alsar Ahmed Hassanein Zinetula Insepov

Growing environmental awareness has renewed interest in thorium as a nuclear fuel, underscoring the need for further studies to evaluate how reactors perform when conventional fuels are replaced with thorium-based alternatives. In this study, thermal hydraulics and solid mechanics computations were simulated using COMSOL multiphysics to investigate the safe operating conditions of a heavy water reactor with thorium-based fuel. The thermo-mechanical analysis of the fuel rod under transient heating conditions provides critical insights into strain, displacement, stress, and coolant flow behavior at elevated volumetric heat sources. After 3 s of heating, the strain distribution in the fuel exhibits a high-strain core surrounded by a low-strain rim, with peak volumetric strain increasing nearly linearly from 0.006 to 0.014 as heat generation rises. Displacement profiles confirm that radial deformation is concentrated at the outer surface, while axial elongation remains uniform and scales systematically with power. The resulting von Mises stress fields show maxima at the outer surface, increasing from ~0.06 to 0.15 GPa at the centerline with higher heat input but remaining within structural safety margins. Cladding simulations demonstrate nearly uniform axial expansion, with displacements increasing from ~0.012 mm to 0.03 mm across the investigated power range, and average strain remains negligible (≈10−4), while mean stresses increase moderately yet stay well below the yield strength of zirconium alloys, confirming safe elastic behavior. Hydrodynamic analysis shows that coolant velocity decreases smoothly along the axial direction but maintains stability, with only minor reductions under increased heat sources. Overall, the coupled thermo-mechanical and fluid-dynamic results confirm that both the fuel and cladding remain structurally stable under the studied conditions. By using COMSOL’s multiphysics capabilities, and unlike most legacy codes optimized for uranium-based fuel, this work is designed to easily incorporate non-traditional fuels such as thorium-based systems, including user-defined material properties, temperature-dependent thermal polynomial formulas, and mechanical response.

Journal of Nuclear Engineering doi: 10.3390/jne6040052

Authors: Harvey Oliver Plows Marat Margulis

Fibre Bragg grating (FBG) grid sensors are an underexplored technology with potential to benefit nuclear thermal hydraulics experiments. This paper presents a new FBG grid sensor consisting of 38 FBGs across 8 flow-crossing chords. Using this sensor, experiments determined for the first time that an FBG grid can detect large air bubbles rising in standing liquids—demonstrated in both columns of water and 20W50 automotive oil. The instrument’s sensitivity was quantified by comparing its measurements to high-speed camera recordings. Analysis of Bragg wavelength shift timings on each chord enabled the surface of a bubble to be reconstructed using the air–oil data. Finally, the increase in Bragg wavelength when bubbles interact with the FBG grid suggests a variant sensing principle different from that reported in the literature for FBG grids in flowing liquids.

Journal of Nuclear Engineering doi: 10.3390/jne6040051

Authors: Glenn A. McRae Christopher E. Coleman

High concentrations of hydrogen isotopes have been observed at the ends of CANDU Zr-2.5Nb pressure tubes in the region associated with the rolled joints with 403 stainless steel end fittings. These concentrations are above current regulatory limits, causing concerns over how long pressure tubes should remain in service. This paper reviews two differing interpretations of the mechanisms for these high concentrations, leading to two conclusions. Ingress after about 30 y is attributed to pressure tube sag creating a crevice between the end fitting and the top of the tube that provides a window for hydrogen isotopes to enter from the annulus gas under reducing conditions. Small additions of oxygen should close this window. A new mechanism is suggested to explain deuteride precipitates past the rolled joint contact region after about 30 y. Surprisingly, the mechanism relies on deuterium and protium diffusing in solution at the same rate, i.e., no mass-dependent isotope effect.

Journal of Nuclear Engineering doi: 10.3390/jne6040050

Authors: Shichang Liu Binji Wang

With the increasing demand for high-fidelity neutronics analysis and the development of computer technology, the Monte Carlo method is becoming increasingly important, especially in the critical analysis of initial core and shielding calculations [...]

Journal of Nuclear Engineering doi: 10.3390/jne6040049

Authors: Takafumi Narukawa Takashi Takata Xiaoyu Zheng Hitoshi Tamaki Yasuteru Sibamoto Yu Maruyama Tsuyoshi Takada

Although dynamic probabilistic risk assessment (PRA) techniques have advanced in their ability to represent the progression of events over time, the formulation of suitable risk importance measures for these methods still poses a substantial challenge. In particular, it is difficult to reflect the full breadth and multidimensional character of the risk information produced by dynamic PRA. In this study, we introduce a set of new importance measures derived from the risk triplet perspective: (i) Timing-Based Worth (TBW), which expresses diversity in scenario occurrence time; (ii) Frequency-Based Worth (FBW), which captures the probability of different scenarios; and (iii) Consequence-Based Worth (CBW), which characterizes scenario consequences. Formal definitions of these three indices are provided, and a conceptual scheme for integrated importance evaluation is proposed to support multidimensional analysis. As an initial demonstration, TBW and FBW are applied to a simplified reliability case using a dynamic PRA framework built on the continuous Markov chain Monte Carlo (CMMC) approach. This application is used to test their interpretability and the internal consistency of the proposed scheme. The findings suggest that TBW and FBW make it possible to conduct more holistic importance evaluations, taking into account resilience effects and temporal diversity in addition to conventional frequency-based perspectives. Such an extension is expected to increase the usefulness of dynamic PRA outputs for risk-informed decision-making.

Journal of Nuclear Engineering doi: 10.3390/jne6040048

Authors: Xinsheng Xu Yiyang Ye Yingjie Wu Yanhua Zheng

The High Temperature Gas-cooled Reactor (HTGR) is a Generation IV advanced nuclear reactor, which can realize inherent safety and prevent core melt. The Institute of Nuclear and New Energy Technology (INET) of Tsinghua University developed a commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble bed Module project (HTR-PM), which entered commercial operation on 6 December 2023. A passive Reactor Cavity Cooling System (RCCS) was designed for HTR-PM to export heat from the reactor cavity during normal operation and also in accident conditions, keeping the safety of the reactor pressure vessel (RPV) and reactor cavity. The RCCS of HTR-PM has been designed as three independent sets; the normal operation of two sets of RCCS can guarantee the safety of the PRV and reactor activity. The heat can be transferred from the RPV to the final heat sink atmosphere through thermal radiation and natural convection in the reactor cavity, and the natural circulation of water and air in the RCCS. The CAVCO code was developed by the INET to simulate the behavior of an RCCS. In this paper, assuming different RPV temperatures and different ambient temperatures, as well as assuming all or parts of the RCCS sets work, the performances of RCCS are studied by CAVCO to evaluate its operational reliability, so as to provide a reference for further optimization. The analysis results indicate that even under hypothetically extremely RPV temperatures, two sets of RCCS could effectively remove heat without causing water boiling or system failure. However, during the winter when ambient temperatures are low, particularly when the reactor operates at a lower RPV temperature, additional attention must be given to the operational safety of the system. It is crucial to prevent system failure caused by the freezing of circulating water and the potential cracking of water-cooling pipes due to freezing. Depending on the reactor status and ambient conditions, one or all three sets of RCCS may need to be taken offline. In addition, the maximum heat removal capacity of the RCCS with only two sets operational exceeds the design requirement of 1.2 MW. When the ambient temperature fluctuates significantly, it may be advisable to increase the number of available RCCS sets to mitigate the effect of abrupt changes in cooling water temperature on pipeline thermal stress.

Journal of Nuclear Engineering doi: 10.3390/jne6040047

Authors: Yuhua Hang Heng Wang Yuanqing Liu Zhen Cai Bin Zhu Jinna Mei Guorui Zhu

As a critical component of nuclear and thermal energy conversion systems, the long-term safe operation of a steam generator depends on the structural integrity of its tube bundles. Foreign objects introduced into the secondary side can induce flow-induced vibrations and wear, potentially causing tube wall damage and unplanned outages, thereby affecting overall system reliability. This study systematically investigates the flow-induced vibration behavior of foreign objects within steam generator tube bundles and explores the influence of object geometry through three-dimensional fluid–structure interaction (FSI) simulations. The foreign objects are modeled as single-degree-of-freedom rigid bodies, and their dynamic responses are captured using a coupled flow–motion framework. Results reveal that object geometry significantly influences flow separation, variations in lift and drag forces, and displacement characteristics. Cylindrical and irregular objects exhibit stable, low-amplitude vibrations; plate-shaped objects experience restricted motion due to large drag areas and symmetric contact constraints; whereas helical objects show the largest displacements arising from coupled axial–radial vibrations and complex vortical structures. These findings demonstrate that the interplay between aerodynamic forces and geometric complexity strongly governs the flow-induced vibration of foreign objects, offering insights into their motion behavior and potential impact on steam generator tube bundle integrity.

Journal of Nuclear Engineering doi: 10.3390/jne6040046

Authors: Nami Yamamoto Mami Kagimoto Yohei Ueno Takafumi Narukawa Takashi Takata

Following the Great East Japan Earthquake in 2011, there has been an increased focus on risk assessment and the practical application of its findings to safety enhancement. In particular, dynamic probabilistic risk assessment (PRA) used in conjunction with plant dynamics analysis is being considered for accident management (AM) and operational support. Determining countermeasure priorities in AM can be challenging due to the diversity of accident scenarios. In multi-unit operations, the complexity of scenarios increases in cases of simultaneous disasters, which makes establishing response operations priorities more difficult. Dynamic PRA methods can efficiently generate and assess complex scenarios by incorporating changes in plant state. This paper introduces the continuous Markov chain Monte Carlo (CMMC) method, a dynamic PRA approach, as a tool for prioritizing countermeasures to support nuclear power plant operations. The proposed method involves three steps: (1) generating exhaustive scenarios that include events, operator actions, and system responses; (2) classifying scenarios according to countermeasure patterns; and (3) assigning priority based on risk data for each pattern. An evaluation was conducted using a simple plant model to analyze event countermeasure patterns for addressing steam generator tube rupture during single-unit operation. The generated scenario patterns included depressurization by opening a pressurizer relief valve (DP), depressurization via heat removal through the steam generator (DSG), and both operations combined (DP + DSG). The timing of the response operations varied randomly, resulting in multiple scenarios. The assessment, based on reactor pressure vessel water level and the potential for core damage, showed that the time margin to core damage depended on the countermeasure pattern. The findings indicate that the effectiveness of each countermeasure can be evaluated and that it is feasible to identify which countermeasure should be prioritized.

Journal of Nuclear Engineering doi: 10.3390/jne6040045

Authors: Mariarosa Giardina Calogera Lombardo

Helically coiled tube heat exchangers (HCT) are recognized as promising solutions for steam generator applications in Small Modular Reactors (SMRs), where compactness and high thermal performance are crucial. The complex geometry of HCTs, however, substantially increases the difficulty of accurately estimating pressure drops, particularly under two-phase flow conditions. Over the last decade, several predictive correlations have been suggested, and their applicability is often limited to specific ranges of geometry and operating pressure. The present study examines correlations proposed during the previous decade, aiming to clarify their applicability limits. Validation is carried out using experimental datasets from the literature, enabling a rigorous evaluation of predictive accuracy, robustness, and generality.

Journal of Nuclear Engineering doi: 10.3390/jne6040044

Authors: Abdo Ez Aldeen Dzianis Litskevich Christopher Grove Seddon Atkinson Anna Detkina Hasnain Gulzar

Compared to traditional large-scale reactors, the more heterogeneous, boron-free SMR cores create additional challenges for accurate multiphysics simulations. Therefore, advanced modelling and simulation tools should be used to create high-fidelity, high-accuracy, and computationally efficient multiphysics and multiscale solvers. These solvers can evaluate the safety and performance of SMRs and could be attractive for industrial applications if the computational power requirements were reasonably low. The first crucial step in building a computationally efficient simulation model is to define an SMR benchmark model. This model is a reference for validating the simulation results. In this paper, the benchmark model is a NuScale-like SMR, where the Serpent code has been utilized to run the neutronic simulation. The neutronic simulation was then performed again in the benchmark model, this time utilizing OpenMC code. The results of the Serpent and OpenMC codes were compared in terms of the reactivity coefficient, control rod worth and radial and axial power distribution. By comparing two different codes to validate the simulation of the NuScale-like benchmark, OpenMC will be utilized for future work, such as generating the nuclear material cross-section data for core simulators.

Journal of Nuclear Engineering doi: 10.3390/jne6040043

Authors: Sumith Yesudasan Mamshad Mohammed Joseph Marcello Mark Taylor

This study presents a molecular dynamics (MD) investigation of water evaporation in copper nanochannels, with a focus on accurately modeling copper–water interactions through forcefield calibration. The TIP4P/2005 water model was coupled with the Modified Embedded Atom Method (MEAM) for copper, and the oxygen–copper Lennard–Jones (LJ) parameters were systematically tuned to match experimentally reported water contact angles (WCAs) on Cu (111) surfaces. Contact angles were extracted from simulation trajectories using a robust five-step protocol involving 2D kernel density estimation, adaptive thresholding, circle fitting, and mean squared error (MSE) validation. The optimized forcefield demonstrated strong agreement with experimental WCA values (50.2°–82.3°), enabling predictive control of wetting behavior by varying ε in the range 0.20–0.28 kcal/mol. Using this validated parameterization, we explored nanoscale evaporation in copper channels under varying thermal loads (300–600 K). The results reveal a clear temperature-dependent transition from interfacial-layer evaporation to bulk-phase vaporization, with evaporation onset and rate governed by the interplay between copper–water adhesion and thermal disruption of hydrogen bonding. These findings provide atomistically resolved insights into wetting and evaporation in metallic nanochannels, offering a calibrated framework for simulating phase-change heat transfer in advanced thermal management systems.

Journal of Nuclear Engineering doi: 10.3390/jne6040042

Authors: Taylor E. Grubbs Igor A. Bolotnov

The two-phase flow within a Boiling Water Reactor steam separator is investigated using an interface capturing method. The simulations are focused on resolving the flow around the first pickoff ring which is the highest contributor to steam carryunder phenomenon. Multiple simulations are conducted of varying levels of resolution to evaluate the capabilities of interface capturing technique for this challenging problem. First, high-resolution simulations of the flow using a simplified 30° wedge are conducted without a swirling velocity field present in the actual system. In order to understand the flow field generated by the separator swirler, secondary simulations of single-phase flow passing through a swirler model are conducted. Using this information, a coarse simulation of the full 360° model was performed, which incorporated the effect of the swirler using a custom inflow boundary condition. Instantaneous carryunder/carryover along with void fraction and film thickness are evaluated at the pickoff ring entrance. Overall, these simulations demonstrate that interface capturing simulations can be an accurate tool for studying full-scale components within nuclear power plants.

Journal of Nuclear Engineering doi: 10.3390/jne6040041

Authors: Hung-Tsung Tsai Bo-Jun Lu Yuh-Ming Ferng Yu Sun

Helical coiled tube (HCT) heat exchangers (HXs) are used in the nuclear industry, particularly in the residual heat removal systems of nuclear power plants (NPPs) and steam generators for small modular reactors. In this study, a single-phase CFD model was developed to investigate non-uniform circumferential distributions in the local wall heat transfer characteristics of a vertical HCT to obtain localized information critical for the safety of NPPs. In a comparison, the predicted circumferential heat transfer characteristics agreed well with the measured data. Governed by centrifugal/gravitational forces, these non-uniform distributions are clearly visible in the results, explaining the test data. We performed additional simulations of the conjugated heat transfer from the hot fluid of the shell side to the cold fluid of the tube side, confirming that the inhomogeneity of circumferential distributions in HCTs is due to the assumption of a constant heat flux boundary condition.

Journal of Nuclear Engineering doi: 10.3390/jne6040040

Authors: Marat Margulis Mustafa J. Bolukbasi

Nuclear energy plays a critical role in global decarbonisation, but its expansion raises concerns about the proliferation risks associated with conventional fuel cycles. This study addresses this challenge by evaluating Am-241 doping as a method to enhance the intrinsic proliferation resistance of nuclear fuel. Using full-core simulations across Pressurised Water Reactors (PWRs), Boiling Water Reactors (BWRs), and Molten Salt Reactors (MSRs), the research assesses the impact of Am-241 on isotopic composition, reactor performance, and safety. The results show that Am-241 reliably increases the Pu-238 fraction in spent fuel above the 6% threshold, which significantly complicates its use in nuclear weapons. Additionally, Am-241 serves as a burnable poison, reducing the need for conventional absorbers without compromising operational margins. Economic modelling indicates that the levelised cost of electricity (LCOE) increases modestly, with the most notable impact observed in MSRs due to continuous doping requirements. The project concludes that Am-241 doping offers a passive, fuel-intrinsic safeguard that complements existing verification regimes. Adoption of this approach may require adjustments to regulatory frameworks, particularly in fuel licencing and fabrication standards, but could ultimately support the secure expansion of nuclear energy in regions with heightened proliferation concerns.

Journal of Nuclear Engineering doi: 10.3390/jne6040039

Authors: Giambattista Guidi Anna Carmela Violante Francesca Romana Macioce

The selection of suitable sites for the disposal of radioactive waste constitutes a critical component of nuclear waste management. This study presents an original methodological proposal based on the Analytic Hierarchy Process (AHP), designed to support early-stage site screening for a near-surface repository in Italy. AHP could be used to identify appropriate locations, focusing on 51 areas that have already undergone a preliminary screening phase. These areas, included in the National Map of Suitable Areas (CNAI), were selected as they fulfill all the technical requirements (geological, geomorphological, and hydraulic stability) necessary to ensure the safety performance of the engineering structures to be implemented through multiple artificial barriers, as specified in Technical Guide N. 29. The proposed methodology is applicable in cases where multiple sites listed in the CNAI have been identified as potential candidates for hosting the repository. A panel of 20 multidisciplinary experts, including engineers, environmental scientists, sociologists, and economists, evaluated two environmental, two economic, and two social criteria not included among the criteria outlined in Technical Guide N. 29. Pairwise comparisons were aggregated using the geometric mean, and consistency ratios (CRs) were calculated to ensure the coherence of expert judgements. Results show that social criteria received the highest overall weight (0.53), in particular the “degree of site acceptability”, followed by environmental (0.28) and economic (0.19) criteria. While the method does not replace detailed site investigations (which will nevertheless be carried out once the site has been chosen), it can facilitate the early identification of promising areas and guide future engagement with local communities. The approach is reproducible, adaptable to additional criteria or national requirements, and may be extended to other countries facing similar nuclear waste management challenges.

Journal of Nuclear Engineering doi: 10.3390/jne6030038

Authors: Hossam H. Abdellatif Joshua Young David Arcilesi Richard Christensen

The molten salt reactor (MSR) is a prominent Generation IV nuclear reactor concept that offers substantial advantages over conventional solid-fueled systems, including enhanced fuel utilization, inherent passive safety features, and significant reductions in long-lived radioactive waste. Central to its safety strategy is a reliance on natural circulation (NC) mechanisms, which eliminate the need for active pumping systems and enhance system reliability during normal and off-normal conditions. However, the challenges associated with molten salts, such as their high melting points, corrosivity, and material compatibility issues, render experimental investigations inherently complex and demanding. Therefore, the use of high-Pr-number surrogate fluids represents a practical alternative for studying molten salt behavior under safer and more accessible experimental conditions. In this study, a single-phase natural circulation loop setup at the University of Idaho’s Thermal–Hydraulics Laboratory was employed to investigate NC behavior under various operating conditions. The RELAP5-3D code was initially validated against water-based experiments before employing Therminol-66, a high-Prandtl-number surrogate for molten salts, in the natural circulation loop for the first time. The RELAP5-3D results demonstrated good agreement with both steady-state and transient experimental results, thereby confirming the code’s ability to model NC behavior in a single-phase flow regime. The results also highlighted certain experimental limitations that should be addressed to enhance the NC loop’s performance. These include increasing the insulation thickness to reduce heat losses, incorporating a dedicated mass flow measurement device for improved accuracy, and replacing the current heater with a higher-capacity unit to enable testing at elevated power levels. By identifying and addressing the main causes of these limitations and uncertainties during water-based experiments, targeted improvements can be implemented in both the RELAP5 model and the experimental setup, thereby ensuring that tests using a surrogate fluid for MSR analyses are conducted with higher accuracy and minimal uncertainty.

Journal of Nuclear Engineering doi: 10.3390/jne6030037

Authors: Karl N. Fleming Bengt O. Y. Lydell Mary Presley Ali Mosleh Wadie Chalgham

Utilities that operate nuclear power plants are increasingly using probabilistic risk assessments (PRAs) to make day-to-day decisions on design, operations, and maintenance and to support risk-informed applications. These applications require high-quality and complete PRAs to ensure that the decisions and proposed changes are technically well-founded. Such PRAs include the modeling and quantification of PRA models for accident sequences initiated by internal floods and high-energy line breaks. To support PRA updates and upgrades for such sequences, the Electric Power Research Institute (EPRI) has sponsored ongoing research to develop and refine guidance and generic data that can be used to estimate initiating event frequencies for internal flood- and high-energy line break-induced accident sequences. In 2023, EPRI published the fifth revision of a generic database for these initiating event frequencies. This revision produced advancements in the methodology for passive component reliability, including the quantification of aging effects on pipe rupture frequencies and the capability to adjust these frequencies to account for enhancements to integrity management strategies associated with leak inspections and non-destructive examinations. The purpose of this paper is to present these enhancements and illustrate their application with selected examples.

Journal of Nuclear Engineering doi: 10.3390/jne6030036

Authors: Ana Ivan Mariano Tarantino Mărioara Abrudeanu Daniela Diaconu Daniela Gugiu

This study reviews the release potential of volatile radionuclides in the ALFRED reactor, providing data for source-term evaluations under both normal and postulated accident conditions. Using empirical Henry’s law relations and radionuclide inventories, the equilibrium partial pressures and maximum gas phase concentrations of activation and fission products were estimated. Results indicate that mercury, cadmium, and tellurium exhibit the highest volatility under normal operation, with more than 99.995% of radionuclides retained in the liquid lead. Polonium, despite its lower volatility, remains a critical safety concern due to its high radiotoxicity. Under elevated temperatures, such as those in an unprotected loss-of-flow (ULOF) scenario, increased release rates for volatile species are expected. In accident conditions involving a defective fuel assembly, fission products, including iodine, caesium, and noble gases, significantly contribute to the gas-phase radiological source term. These findings confirm the essential role of continuous cover gas monitoring and efficient purification systems in maintaining reactor safety.

Journal of Nuclear Engineering doi: 10.3390/jne6030035

Authors: Shusuke Hamaguchi Takafumi Narukawa Takashi Takata

In probabilistic risk assessment (PRA), the fracture limit of fuel cladding tubes under loss-of-coolant accident conditions plays a critical role in determining the core damage, highlighting the need for accurate modeling of cladding tube fracture behavior. However, for high-burnup cladding tubes, it is often infeasible to conduct extensive experiments due to limited material availability, high costs, and technical constraints. These limitations make it difficult to acquire sufficient data, leading to substantial epistemic uncertainty in fracture modeling. To enhance the realism of PRA results under such constraints, it is essential to develop methods that can effectively reduce epistemic uncertainty using limited experimental data. In this study, we propose a Bayesian approach for designing experimental conditions based on a widely applicable information criterion (WAIC) in order to effectively reduce the uncertainty in the prediction of fuel cladding tube fracture with limited data. We conduct numerical experiments to evaluate the effectiveness of the proposed method in comparison with conventional approaches based on empirical loss and functional variance. Two cases are considered: one where the true and predictive models share the same mathematical structure (Case 1) and one where they differ (Case 2). In Case 1, the empirical loss-based design performs best when the number of added data points is fewer than approximately 10. In Case 2, the WAIC-based design consistently achieves the lowest Bayes generalization loss, demonstrating superior robustness in situations where the true model is unknown. These results indicate that the proposed method enables more informative experimental designs on average and contributes to the effective reduction in epistemic uncertainty in practical applications.

Journal of Nuclear Engineering doi: 10.3390/jne6030034

Authors: Marino Corrado Francesca Crivelli Silvio Cao Laura Savoldi

The HYPEX® process is a novel method for conditioning spent ion exchange resins from nuclear power plants, aiming to reduce final waste volume and carbon emissions by stabilizing the resins in metakaolin-based geopolymers. This study addresses the challenge posed by the natural variability of commercial metakaolin and defines a testing strategy to ensure consistent performance of the final matrix. The reactivity of two batches of metakaolin, characterized by comparable chemical composition and BET surface area, was evaluated by monitoring temperature evolution during geopolymerization at varying water-to-solid ratios. The resulting geopolymers were tested for compressive strength, water permeability, and strontium leachability to assess correlations between precursor properties and final matrix performance. Despite similar compositions, the two batches showed marked differences in compressive strength that could be linked to early thermal behavior. These findings demonstrate that conventional precursor characterization is insufficient to guarantee reproducibility and that thermal profiling is useful to predict mechanical performance. The results suggest the implementation of thermal response monitoring as a quality control tool to ensure the reliability of geopolymer wasteforms in nuclear applications. A simplified analytical model for the thermal evolution during geopolymerization was also developed, matching qualitatively the measured evolution, to suggest scale-up rules from laboratory specimens to full-scale drums, which should be achieved while preserving the thermal evolution.

Journal of Nuclear Engineering doi: 10.3390/jne6030033

Authors: Andrew Politz Paolo Tancioni Oskar Searfus Eric Aboud Kelly Jordan Daniel Siefman

This work investigates a 4He-detector active interrogation system that leverages neutron spectroscopy to classify nuclear waste streams. MCNP models tested the concept through the simulation of a D-D neutron generator, an array of 4He detectors, and various waste compositions. The fast-neutron Differential Die-Away signature was augmented with a neutron-energy discrimination signature. This signature isolates induced fission neutrons, the energy of which is greater than that of the D-D monoenergetic spectrum. With the incorporation of this spectroscopic technique, the measurement time decreased by 3–9% (depending on the degree of neutron moderation and absorption presented by the sample), demonstrating how neutron spectroscopy can enhance active interrogation methods. The reduced measurement times would have significant financial and logistical benefits for facilities with large footprints of low-level waste production.

Journal of Nuclear Engineering doi: 10.3390/jne6030032

Authors: Nahum Contreras-Pérez Heriberto Sánchez-Mora Sergio Quezada-García Armando Miguel Gómez Torres Ricardo Isaac Cázares Ramírez

Hydrogen generation and the correct simulation of severe accidents have been of utmost importance since the Fukushima Dai-ichi accident. QUENCH experiments are quite useful for validating mathematical models implemented in system codes for early-phase severe accidents, where hydrogen generation, fuel rod temperature, and their deterioration during these conditions are of vital importance. This paper presents a new system code, TLANESY, designed for the simulation of thermal–hydraulic systems with two-phase flow (mainly water) and with application in the analysis of severe accidents during the early phase. The computational implementation consists of fast-running numerical methods and their validation with experimental data from the QUENCH-01 experiment. The results showed an error with respect to the total hydrogen generation of approximately 0.6%. A stand-alone sensitivity analysis was also performed with some parameters related to the cladding, where it was shown that variation in the thermal conductivity by 15% can alter the total hydrogen generation by up to 5%, indicating that impurities in this material can have a significant impact on this Figure of Merit.

Journal of Nuclear Engineering doi: 10.3390/jne6030031

Authors: Sai Kiran Kokkiligadda Cathleen Barker Emily Gunger Jalen Johnson Brice Turner Andreas Enqvist

The capability to discriminate among nuclear fuel properties is essential for a successful nuclear safeguard and security program. Accurate nuclear material identification is hindered due to challenges such as differing levels of enrichments, weak radiation signals in the case of fresh nuclear fuel, and complex self-shielding effects. This study explores the application of supervised machine learning algorithms to digitized radiation detector data for classifying signatures of special nuclear materials. Three scintillation detectors, an EJ-309 liquid scintillator, a CLYC crystal scintillator, and an EJ-276 plastic scintillator, were used to measure gamma-ray and neutron data from special nuclear material at the National Criticality Experiments Research Center (NCERC) at the National Nuclear Security Site (NNSS), at Nevada, USA. Radiation detector pulse data was extracted from the collected digitized data and applied to three separate supervised learning models: Random Forest, XGBoost, and a feedforward Deep Neural Network, chosen for their wide-spread use and distinct data ingest and processing analytics. Through model refinement, such as adding an additional parameter feature, an accuracy of greater than 95% was achieved. Analysis on model parameter feature importance revealed Countrate, which is the overall gamma-ray and neutron incidents for each detector, was the most influential parameter and essential to include for improved classification. Initial model versions not including the Countrate parameter feature failed to classify. Supervised learning models allow for measured gamma-ray and neutron pulse data to be used to develop effective identification and discrimination between material compositions of different fuel assemblies. The study demonstrated that traditional pulse shape parameters alone were insufficient for discriminating between special nuclear materials; the addition of Countrate markedly improved model accuracy but all models were heavily dependent on this specific feature, thus illustrating the need for alternative, more distinct parameter features. The machine learning development framework captured in this study will be beneficial for future applications in discriminating between different fuel enrichments and additives such as burnable poisons.

Journal of Nuclear Engineering doi: 10.3390/jne6030030

Authors: Tyler Law Tracey Spoerer Steven Reese

The results of a recently developed Serpent 2 model of the Oregon State TRIGA® Reactor (OSTR) are compared to the results from the OSTR MCNP® model and measured values for reactor steady state behavior. This benchmark comparison is performed using fresh fuel isotopic data and measured reactivity values at the beginning of the current core life in 2008 to negate burnup uncertainties in calculated values. Reactivity bias, integral control rod reactivity worths, core excess reactivity, shutdown margin, the fuel temperature coefficient of reactivity, and kinetic parameters calculated by Serpent 2 and MCNP® are compared to the measured values. The results from the Serpent 2 model strongly agree with both MCNP® results and measured values and are within one standard deviation of each other in all cases, with the exception of the Serpent 2 calculated total control rod reactivity worth, which slightly under-predicts the total rod worth when compared to the measured value despite the MCNP® and Serpent 2 calculated total rod worth values being within each other’s 1σ standard deviations.

Journal of Nuclear Engineering doi: 10.3390/jne6030029

Authors: Alistair F. Holdsworth Edmund Ireland Harry Eccles

Some of nuclear power’s primary detractors are the unique environmental challenges and impacts of radioactive wastes generated during fuel cycle operations. Key benefits of spent fuel reprocessing (SFR) are reductions in primary high active waste (HAW) masses, volumes, and lengths of radiotoxicity at the expense of secondary waste generation and high capital and operational costs. By employing advanced waste management and resource recovery concepts in SFR beyond the existing standard PUREX process, such as minor actinide and fission product partitioning, these challenges could be mitigated, alongside further reductions in HAW volumes, masses, and duration of radiotoxicity. This work assesses various current and proposed SFR and fuel cycle options as base cases, with further options for fission product partitioning of the high heat radionuclides (HHRs), rare earths, and platinum group metals investigated. A focus on primary waste outputs and the additional energy that could be generated by the reprocessing of high-burnup PWR fuel from Gen III(+) reactors using a simple fuel cycle model is used; the effects of 5- and 10-year spent fuel cooling times before reprocessing are explored. We demonstrate that longer cooling times are preferable in all cases except where short-lived isotope recovery may be desired, and that the partitioning of high-heat fission products (Cs and Sr) could allow for the reclassification of traditional raffinates to intermediate level waste. Highly active waste volume reductions approaching 50% vs. PUREX raffinate could be achieved in single-target partitioning of the inactive and low-activity rare earth elements, and the need for geological disposal could potentially be mitigated completely if HHRs are separated and utilised.

Journal of Nuclear Engineering doi: 10.3390/jne6030028

Authors: Nicholas Herring Robert Salko Mehdi Asgari

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type.

Journal of Nuclear Engineering doi: 10.3390/jne6030027

Authors: Ben Impson Mohamed Elhareef Zeyun Wu Braden Goddard

Pebble Bed Reactors (PBRs) represent a new generation of nuclear reactors. However, modeling TRi-structural ISOtropic (TRISO) fuel particles employed in PBRs presents a unique challenge in comparison to most conventional reactor designs. Rapid generation of different possible fuel particle configurations for Monte-Carlo simulations provides improved insights into the effects of particle distribution irregularities on the neutron economy. Defective pebbles could cause changes in the neutron flux in a nuclear reactor due to increased or decreased moderating effects. Different configurations of particle fuel also impact isotope production within the nuclear reactor. This study simulates several TRISO configurations representing limited capabilities of randomization algorithms, manufacturing defects configurations and/or special pebble design. All predictions are compared to an equivalent homogenized model used as baseline. The results show that the TRISO configuration has a non-negligible impact on the parameters under consideration. To explain these results, the ratio of the thermal flux of each model to the thermal flux of the homogeneous model is calculated. A clear pattern is observed in the data: as irregularities in the moderator medium emerge due to the distribution of TRISO particles, the neutron spectrum softens, leading to higher values of k∞ and better fuel utilization. This dependence of the spectrum on the TRISO configuration is used to explain the pattern observed in the depletion calculation. The results open the possibility of optimizing the TRISO configuration in manufactured pebbles for fuel utilization and safeguards. Future work should focus on full core simulations to determine the extent of these findings.

Journal of Nuclear Engineering doi: 10.3390/jne6030026

Authors: Valerio Giusti

Burnable poisoning and fuel enrichment zoning are two techniques often combined in order to optimize the fuel assembly behavior during the burnup cycle. In the present work, these two techniques will be applied to the 2D optimization of the fuel assembly conceptual design for the supercritical water-cooled reactor developed in the framework of the Joint European Canadian Chinese development of Small Modular Reactor Technology project, funded within the Euratom Research and Training programme 2019–2020. The initial configuration of the fuel assembly does not include any burnable absorbers and uses a homogeneous fuel enrichment of 7.5% in 235U. The infinite multiplication factor, k∞, starts from approximately 1.32 and drops, almost linearly, to 1.0 after a burnup of 40.0 MWd·kg−1. The uniform enrichment is, however, responsible for a pin-power peaking factor that with fresh fuel starts from 1.32 and reduces to 1.08 at the end of the burnup cycle. A simplified analytical model is developed to assess the effect of different lumped burnable absorbers on the time dependence of the assembly k∞. It is shown that using an adequate number of B4C rods, positioned in the outer wall of the fuel assembly, together with a suitable distribution of six different 235U enrichments, it allows for obtaining an assembly k∞ factor that starts from 1.11 at the beginning of the cycle and remains quite constant over a large fraction of the burnup cycle. Moreover, the pin-power peaking factor is reduced to 1.03 at the beginning of the cycle and remains almost unchanged until the end of the burnup cycle.

Journal of Nuclear Engineering doi: 10.3390/jne6030025

Authors: Mauricio Gago Steffen Antusch Alexander Klein Arkadi Kreter Christian Linsmeier Michael Rieth Bernhard Unterberg Marius Wirtz

Powder injection molding (PIM) has been used to produce nearly net-shaped samples of tungsten-based alloys. These alloys have been previously shown to have favorable characteristics when compared with standard ITER-grade tungsten. Six different alloys were produced with this method: W-1TiC, W-2Y2O3, W-3Re-1TiC, W-3Re-2Y2O3, W-1HfC and W-1La2O3-1TiC. These were tested alongside ITER-grade tungsten in the PSI-2 linear plasma device under ITER-relevant plasma and heat loads to assess their suitability for use in a fusion reactor. All materials showed good behavior when exposed to the lower pulse number tests (≤1000 ELM-like pulses), although standard tungsten performed slightly better, with no observable difference in surface roughness. High-power shots, namely one laser pulse of 1.6 GWm−2, revealed that samples containing yttria are more prone to melting and droplet ejection. After high pulse number tests (10,000 and 100,000 pulses), with and without plasma, the reference tungsten showed the most cracking and highest surface roughness of all materials, while the PIM samples seemed to have a higher resistance to cracking. This can be attributed to the higher ductility of these alloys, particularly those containing rhenium. This means that tungsten-based alloys, whether produced via PIM or other methods, could potentially be used in certain areas of a fusion reactor.

Journal of Nuclear Engineering doi: 10.3390/jne6030024

Authors: Dan Gabriel Cacuci

This work presents the mathematical frameworks of the “First-Order Features Adjoint Sensitivity Analysis Methodology for Neural Integro-Differential Equations of Volterra-Type” (1st-FASAM-NIDE-V) and the “Second-Order Features Adjoint Sensitivity Analysis Methodology for Neural Integro-Differential Equations of Volterra-Type” (2nd-FASAM-NIDE-V). It is shown that the 1st-FASAM-NIDE-V methodology enables the efficient computation of exactly-determined first-order sensitivities of the decoder response with respect to the optimized NIDE-V parameters, requiring a single “large-scale” computation for solving the 1st-Level Adjoint Sensitivity System (1st-LASS), regardless of the number of weights/parameters underlying the NIE-net. The 2nd-FASAM-NIDE-V methodology enables the computation, with unparalleled efficiency, of the second-order sensitivities of decoder responses with respect to the optimized/trained weights involved in the NIDE-V’s decoder, hidden layers, and encoder, requiring only as many “large-scale” computations as there are non-zero first-order sensitivities with respect to the feature functions. These characteristics of the 1st-FASAM-NIDE-V and 2nd-FASAM-NIDE-V are illustrated by considering a nonlinear heat conduction model that admits analytical solutions, enabling the exact verification of the expressions obtained for the first- and second-order sensitivities of NIDE-V decoder responses with respect to the model’s functions of parameters (weights) that characterize the heat conduction model.

Journal of Nuclear Engineering doi: 10.3390/jne6030023

Authors: Lanxin Gong Changhong Peng Qingyu Huang

Recent advances in high-fidelity modeling, numerical computing, and data science have spurred interest in model-data integration for nuclear reactor applications. While machine learning often prioritizes data-driven predictions, this study focuses on data assimilation (DA) to synergize physical models with measured data, aiming to enhance predictive accuracy and reduce uncertainties. We implemented deterministic DA methods—Kalman filter (KF) and three-dimensional variational (3D-VAR)—in a one-dimensional single-phase natural circulation loop and extended 3D-VAR to RELAP5, a system code for two-phase loop analysis. Unlike surrogate-based or model-reduction strategies, our approach leverages full-model propagation without relying on computationally intensive sampling. The results demonstrate that KF and 3D-VAR exhibit robustness against varied noise types, intensities, and distributions, achieving significant uncertainty reduction in state variables and parameter estimation. The framework’s adaptability is further validated under oceanic conditions, suggesting its potential to augment baseline models beyond conventional extrapolation boundaries. These findings highlight DA’s capacity to improve model calibration, safety margin quantification, and reactor field reconstruction. By integrating high-fidelity simulations with real-world data corrections, the study establishes a scalable pathway to enhance the reliability of nuclear system predictions, emphasizing DA’s role in bridging theoretical models and operational demands without compromising computational efficiency.

Journal of Nuclear Engineering doi: 10.3390/jne6030022

Authors: Salvatore Almaviva Lidia Baiamonte Marco Pistilli

In current and future fusion devices, detecting hydrogen isotopes, particularly tritium and deuterium, implanted or redeposited on the surface of Plasma-Facing Components (PFCs) will be increasingly important to ensure safe machine operations. The Laser-Induced Breakdown Spectroscopy (LIBS) technique has proven capable of performing this task directly in situ, without handling or removing PFCs, thus limiting analysis times and increasing the machine’s duty cycle. To increase sensitivity and the ability to discriminate between isotopes, LIBS analysis can be performed under different background gases at atmospheric pressure, such as air, He, and Ar. In this work, we present the results obtained on tungsten coatings enriched with deuterium and/or hydrogen as a deuterium–tritium nuclear fuel simulant, measured with the LIBS technique in air, He, and Ar at atmospheric pressure, and discuss the pros and cons of their use. The results obtained demonstrate that both He and Ar can improve the LIBS signal resolution of the hydrogen isotopes compared to air. However, using Ar has the additional advantage that the same procedure can also be used to detect He implanted in PFCs as a product of fusion reactions without any interference. Finally, the LIBS signal in an Ar atmosphere increases in terms of the signal-to-noise ratio (SNR), enabling the use of less energetic laser pulses to improve performance in depth profiling analyses.

Journal of Nuclear Engineering doi: 10.3390/jne6030021

Authors: Xiaoyu Zheng Hitoshi Tamaki Yasuteru Sibamoto Yu Maruyama Tsuyoshi Takada Takafumi Narukawa Takashi Takata

While traditional risk importance measures in probabilistic risk assessment are effective for ranking safety-significant components, they often overlook critical aspects such as the timing of accident progression and consequences. Dynamic probabilistic risk assessment offers a framework to quantify such risk information, but standardized approaches for estimating risk importance measures remain underdeveloped. This study addresses this gap by: (1) reviewing traditional risk importance measures and their regulatory applications, highlighting their limitations, and introducing newly proposed risk-triplet-based risk importance measures, consisting of timing-based worth, frequency-based worth, and consequence-based worth; (2) conducting a case study of Level 2 dynamic probabilistic risk assessment using the Japan Atomic Energy Agency’s RAPID tool coupled with the severe accident code of MELCOR 2.2 to simulate a station blackout scenario in a boiling water reactor, generating probabilistically sampled sequences with quantified timing, frequency, and consequence of source term release; (3) demonstrating that the new risk importance measures provide differentiated insights into risk significance, enabling multidimensional prioritization of systems and mitigation strategies; for example, the timing-based worth quantifies the delay effect of mitigation systems, and the consequence-based worth evaluates consequence-mitigating potential. This study underscores the potential of dynamic probabilistic risk assessment and risk-triplet-based risk importance measures to support risk-informed and performance-based regulatory decision-making, particularly in contexts where the timing and severity of accident consequences are critical.

Journal of Nuclear Engineering doi: 10.3390/jne6030020

Authors: Derar Altarawneh Roman Höllwieser

The study of scalar field theories like the chameleon field model is of increasing interest due to the Universe’s accelerated expansion, which is believed to be caused in part by dark energy. These fields can elude experimental bounds set on them in high-density environments since they interact with matter in a density-dependent way. This paper analyzes the effect of chameleon fields on the quantum gravitational states of ultra-cold neutrons (UCNs) in qBounce experiments with mirrors. We discuss the deformation of the neutron wave function due to chameleon interactions and quantum systems in potential wells from gravitational forces and chameleon fields. Unlike other works that aim to put bounds on the chameleon field parameters, this work focuses on the quantum mechanics of the chameleonic neutron. The results deepen our understanding of the interplay between quantum states and modified gravity, as well as fundamental physics experiments carried out in the laboratory.

Journal of Nuclear Engineering doi: 10.3390/jne6020019

Authors: Nobuyuki Fujisawa

The subject of the effect of pipeline layout on wall thinning in Mihama nuclear power plants was reviewed in relation to flow-accelerated corrosion (FAC). The pipeline consists of a complex layout with a straight pipe, elbow, curved pipe, orifice, and T-junction. To understand the mechanism of wall thinning in the pipeline, the basics of FAC, experimental and numerical approaches, and flow and mass transfer studies of the pipeline were reviewed and compared with actual Mihama pipeline data. The results indicate that the wall thinning in the Mihama pipeline was caused by the asymmetric mass transfer phenomenon arising from the pipeline layout effect induced by the swirl flow, resulting in the generation of a spiral flow downstream of the elbow and an increased mass transfer coefficient downstream of the orifice. Swirl flow can be generated by the coupled T-junction and elbow in the upstream pipeline. Furthermore, related topics in flow and mass transfer studies on short elbows and dual and triple elbows were reviewed in relation to wall thinning, which could depend on the elbow curvature, Reynolds number, and surface roughness. The low-frequency flow oscillation in a short elbow, the swirl flow generation in dual and triple elbows, and the influence of wall roughness could be other sources of the increased mass transfer coefficient in the pipeline.

Journal of Nuclear Engineering doi: 10.3390/jne6020018

Authors: Saikat Basak Lixuan Lu

This research explores Dynamic Probabilistic Risk Assessment (DPRA) using EMRALD to evaluate the reliability and safety of passive safety systems in nuclear reactors, with a focus on mitigating Loss of Coolant Accidents (LOCAs). The BWRX-300 Small Modular Reactor (SMR) is used as an example to illustrate the proposed DPRA methodology, which is broadly applicable for enhancing traditional Probabilistic Safety Assessment (PSA). Unlike static PSA, DPRA incorporates time-dependent interactions and system dynamics, allowing for a more realistic assessment of accident progression. EMRALD enables the modelling of system failures and interactions in real time using dynamic event trees and Monte Carlo simulations. This study identifies critical vulnerabilities in passive safety systems and quantifies the Core Damage Frequency (CDF) under LOCA scenarios. The findings demonstrate the advantages of DPRA over traditional PSA in capturing complex failure mechanisms and providing a more comprehensive and accurate risk assessment. The insights gained from this research contribute to improving passive safety system designs and enhancing nuclear reactor safety strategies for next-generation reactors.

Journal of Nuclear Engineering doi: 10.3390/jne6020017

Authors: Md Shakhawat Hossen Bhuiyan Jintana Meesungnoen Jean-Paul Jay-Gerin

(1) Background: Generation IV supercritical water-cooled reactors (SCWRs), including small modular reactor (SCW-SMR) variants, are pivotal in nuclear technology. Operating at 300–500 °C and 25 MPa, these reactors require detailed understanding of radiation chemistry and transient species to optimize water chemistry, reduce corrosion, and enhance safety. Boron, widely used as a neutron absorber, plays a significant role in reactor performance and safety. This study focuses on the yields of radiolytic species in subcritical and supercritical water exposed to 4He and 7Li recoil ions from the 10B(n,α)7Li fission reaction in SCWR/SCW-SMR environments. (2) Methods: We use Monte Carlo track chemistry simulations to calculate yields (G values) of primary radicals (e−aq, H•, and •OH) and molecular species (H2 and H2O2) from water radiolysis by α-particles and Li3⁺ recoils across 1 picosecond to 0.1 millisecond timescales. (3) Results: Simulations show substantially lower radical yields, notably e−aq and •OH, alongside higher molecular product yields compared to low linear energy transfer (LET) radiation, underscoring the high-LET nature of 10B(n,α)7Li recoil nuclei. Key changes include elevated G(•OH) and G(H2), and a decrease in G(H•), primarily driven during the homogeneous chemical stage of radiolysis by the reaction H• + H2O → •OH + H2. This reaction significantly contributes to H2 production, potentially reducing the need for added hydrogen in coolant water to mitigate oxidizing species. In supercritical conditions, low G(H₂O₂) suggests that H2O2 is unlikely to be a major contributor to material oxidation. (4) Conclusions: The 10B(n,α)7Li reaction’s yield estimates could significantly impact coolant chemistry strategies in SCWRs and SCW-SMRs. Understanding radiolytic behavior in these conditions aids in refining reactor models and coolant chemistry to minimize corrosion and radiolytic damage. Future experiments are needed to validate these predictions.

Journal of Nuclear Engineering doi: 10.3390/jne6020016

Authors: Uku Andreas Reigo Cansu Özcan Kılcan Alan H. Tkaczyk

Supplementary cementitious materials (SCMs) may be prepared using industrial byproduct streams, aiding in the development of a more environmentally sustainable circular economy. However, these byproducts may carry a risk of exhibiting elevated levels of radioactivity because of the preceding processing that may have concentrated the radionuclides naturally occurring in the raw material. This processing causes the byproducts to be considered technologically enhanced naturally occurring radioactive material (NORM). Thus, the safe use of such SCMs requires robust data on the activity concentrations of three main radionuclides (226Ra, 232Th, 40K) represented by the activity concentration index (ACI) used as a radiological suitability indicator. In this work, candidate SCMs derived from the alumina industry byproduct processed bauxite (PB), also referred to as bauxite residue, were assessed by measuring the activity of all available samples, including input raw materials and intermediate substances, through gamma spectrometry. PB was found to significantly impact the final ACI value of the building material. As a key analysis outcome applicable to the substances assessed in this work, no additional dose assessment is required, given the low ACI value of the building materials. This result indicates that, from a radiological perspective, the PB samples studied are suitable precursors for SCMs. In addition, a generalized approach was found to provide good estimations of the ACI value of building materials, which is useful to screen materials for regulatory compliance, without needing to prepare samples of the materials in question.

Journal of Nuclear Engineering doi: 10.3390/jne6020015

Authors: Allan Xi Chen Benjamin F. Sigal John Martinis Alfred YiuFai Wong Alexander Gunn Matthew Salazar Nawar Abdalla Kai-Jian Xiao

We present the performance and application of a commercial off-the-shelf Si PIN diode (Hamamatsu S14605) as a charged particle detector in a compact ion beam system (IBS) capable of generating D–D and p–B fusion charged particles. This detector is inexpensive, widely available, and operates in photoconductive mode under a reverse bias voltage of 12 V, supplied by an A23 battery. A charge-sensitive preamplifier (CSP) is mounted on the backside of the detector’s four-layer PCB and powered by two ±3 V lithium batteries (A123). Both the detector and CSP are housed together on the vacuum side of the IBS, facing the fusion target. The system employs a CF-2.75-flanged DB-9 connector feedthrough to supply the signal, bias voltage, and rail voltages. To mitigate the high sensitivity of the detector to optical light, a thin aluminum foil assembly is used to block optical emissions from the ion beam and target. Charged particles generate step responses at the preamplifier output, with pulse rise times in the order of 0.2 to 0.3 µs. These signals are recorded using a custom-built data acquisition unit, which features an optical fiber data link to ensure the electrical isolation of the detector electronics. Subsequent digital signal processing is employed to optimally shape the pulses using a CR-RCn filter to produce Gaussian-shaped signals, enabling the accurate extraction of energy information. Performance results indicate that the detector’s baseline RMS ripple noise can be as low as 0.24 mV. Under actual laboratory conditions, the estimated signal-to-noise ratios (S/N) for charged particles from D–D fusion—protons, tritons, and helions—are approximately 225, 75, and 41, respectively.

Journal of Nuclear Engineering doi: 10.3390/jne6020014

Authors: Aldo Dall’Osso

The need to model the effect of the assembly environment on the neutronic data has been felt since Smith’s topical article on assembly homogenization techniques. Indeed, simply homogenizing the cross sections using the spatial distribution and energy spectrum of the neutron flux calculated in a single assembly with reflective boundary conditions, neglecting the effect of the proximity of other types of assemblies, can induce inaccuracies affecting the results of core calculations. Many approaches have been proposed to take into account the real environment of the assembly. The purpose of this article is to review these methods to allow the reader to compare them.

Journal of Nuclear Engineering doi: 10.3390/jne6020013

Authors: Silvano Tosti Luca Farina

The exploitation of fusion energy in tokamak reactors relies on efficient and reliable tritium management. The tritium needed to sustain the deuterium–tritium fusion reaction is produced in the Li-based blanket surrounding the plasma chamber, and, therefore, the effective extraction and purification of the tritium bred in the Li-blankets is needed to guarantee the tritium self-sufficiency of future fusion plants. This work introduces a new technology for the extraction of tritium from the Pb–Li eutectic alloy used in liquid blankets. Process units based on the concept of Membrane Gas–Liquid Contactor (MGLC) have been studied for the extraction of tritium from the Pb–Li in the Water Cooled Lithium Lead blankets of the DEMO reactor. MGLC units have been preliminarily designed and then compared in terms of the permeation areas and sizes with the tritium extraction technologies presently under study, namely the Permeator Against Vacuum (PAV) and the Gas–Liquid Contactors (GLCs). The results of this study show that the DEMO WCLL tritium extraction systems using MGLC require smaller permeation areas and quicker permeation kinetics than those based on PAV (Permeator Against Vacuum) devices. Accordingly, the MGLC extraction unit exhibits volumes smaller than those of both PAV and GLC.

Journal of Nuclear Engineering doi: 10.3390/jne6020012

Authors: Ghada Yassin Erik Pönitz Nina Maria Huittinen Dieter Schild Jörg Konheiser Katharina Müller Astrid Barkleit

This study presents a comprehensive analysis of the microstructural characteristics and chemical composition of base and weld materials from reactor pressure vessels in the first (units 1 and 2) and second (unit 8) generations of Russian VVER 440 reactors at the Greifswald nuclear power plant. We measured the specific activities of 60Co and 14C in activated samples from units 1 and 2. 60Co, with its shorter half-life (t1/2 = 5.27 a), is a key dose-contributing radionuclide during decommissioning, while 14C (t1/2 = 5700 a) plays an important role in a geological repository for low- and intermediate-level radioactive waste. Our findings reveal differences in the proportions of trace elements between the base and weld materials as well as between the two reactor generations. Microstructural analysis identified Mo-rich precipitates and (Mn, S)-rich inclusions containing secondary micro-inclusions in the unit 1 and 2 samples. Raman spectroscopy confirmed iron oxides (γ-Fe2O3, Fe3O4), silicates (Mn-SiO3), and Cr2O3/NiCr2O4 in the base metal as well as MnFe2O3 in the weld metal. X-ray photoelectron spectroscopy identified Mn inclusions as MnS, MnS2, or mixed Mn, Fe sulfides, and the Mo precipitates as MoSi2. These findings offer valuable insights into the speciation of elements and the potential release of radionuclides through corrosion processes under repository conditions.

Journal of Nuclear Engineering doi: 10.3390/jne6020011

Authors: Peng Hong Liem

In this study, a 56-group covariance library was generated based on the recently released JENDL-5 covariance data, which cover 105 isotopes. The AMPX-6 code system facilitated the generation of this library. Subsequently, the TSUNAMI-IP code was employed to estimate the uncertainty in the effective neutron multiplication factor (keff) for the critical experiment conducted in the Japanese High-Temperature Test Reactor (HTTR). Our analysis involved comparing results obtained from three nuclear data libraries: JENDL-5, ENDF/B-VIII.0, and ENDF/B-VII.1. The keff uncertainty originated from the nuclear data of JENDL-5, ENDF/B-VIII.0, and ENDF/B-VII.1 and were estimated to be 0.387%, 0.581%, and 0.556%, respectively. Interestingly, when the JENDL-5 covariance library was combined with ENDF/B-VIII.0 for JENDL-5 nuclides lacking covariance data, the keff uncertainty increased to 0.464%. The primary contributors to the keff uncertainty, ranked in decreasing order, were U-235 (nubar), C-12 (n,gamma), U-235 (fission), C-12 (elastic), and U-238 (n,gamma). Notably, significant differences in the keff uncertainty were observed between JENDL-5 and ENDF/B-VIII.0, particularly for U-235 (nubar) and C-12 (elastic). Additionally, the sensitivity coefficients, similarity, and kinetics parameters were evaluated across the three libraries, leading to insightful inter-library comparison results.

Journal of Nuclear Engineering doi: 10.3390/jne6020010

Authors: Jeremy Marquardt Leonard Lucas Stylianos Chatzidakis

Safer and more efficient characterization of radioactive environments requires exploring intelligently, utilizing robotic systems which use smart strategies and physics-based statistical models. Bayesian Optimization (BO) provides one such statistical framework to explainably find the global maximum within noisy contexts while also minimizing the number of trials. For radiation survey and source location, the aid of such a machine learning algorithm could significantly cut down on time and health risks required for maintenance and emergency response scenarios. Maintaining the explainability while increasing the efficiency of the search has been found possible by including the high uncertainty data that is picked up while the agent is in transit. Now that the paths of transit matter to data acquisition they could be optimized as well. This paper introduces reinforcement learning (RL) to the BO search framework. The behavior of this RL additive is observed in simulation over three different datasets of real radiation data. It is shown that the RL additive can cause significant increases to the score of the maximum point discovered, but the computational time cost is increased by nearly 100% while the reconstructed radiation field root mean square error (RMSE) of the BO+RL algorithm matches BO performance within 1%.

Journal of Nuclear Engineering doi: 10.3390/jne6020009

Authors: Maksim V. Bulavin Ivan L. Litvak

An experimental study was conducted on the feasibility of using frozen beads with the properties of Rupert’s drops—solid frozen beads with enhanced strength made from a mixture of aromatic hydrocarbons—in cryogenic neutron moderators utilizing bead technology. It is demonstrated that the use of a new modification of the dosing device with a high discharge rate (approximately 6 units/s) significantly improves process efficiency. With standard pneumatic transport parameters maintained, it was possible to load solid frozen beads made from a mixture of mesitylene and meta-xylene into the cryogenic moderator chamber. The loading speed increased five-fold, while the beads remained intact during pneumatic transport.

Journal of Nuclear Engineering doi: 10.3390/jne6020008

Authors: Dan Gabriel Cacuci

This work presents the general mathematical frameworks of the “First and Second-Order Features Adjoint Sensitivity Analysis Methodology for Neural Integral Equations of Volterra Type” designated as the 1st-FASAM-NIE-V and the 2nd-FASAM-NIE-V methodologies, respectively. Using a single large-scale (adjoint) computation, the 1st-FASAM-NIE-V enables the most efficient computation of the exact expressions of all first-order sensitivities of the decoder response to the feature functions and also with respect to the optimal values of the NIE-net’s parameters/weights after the respective NIE-Volterra-net was optimized to represent the underlying physical system. The computation of all second-order sensitivities with respect to the feature functions using the 2nd-FASAM-NIE-V requires as many large-scale computations as there are first-order sensitivities of the decoder response with respect to the feature functions. Subsequently, the second-order sensitivities of the decoder response with respect to the primary model parameters are obtained trivially by applying the “chain-rule of differentiation” to the second-order sensitivities with respect to the feature functions. The application of the 1st-FASAM-NIE-V and the 2nd-FASAM-NIE-V methodologies is illustrated by using a well-known model for neutron slowing down in a homogeneous hydrogenous medium, which yields tractable closed-form exact explicit expressions for all quantities of interest, including the various adjoint sensitivity functions and first- and second-order sensitivities of the decoder response with respect to all feature functions and also primary model parameters.

Journal of Nuclear Engineering doi: 10.3390/jne6010007

Authors: Alexander Ustinov Jovan Mitrovic Dmitry Ustinov

A lot of novel surface treatment technologies have appeared over the last few decades, offering great possibilities for practical use. Modified surfaces have confirmed their successful application in thermal engineering for boiling heat transfer enhancement and single-phase convection. Several classification approaches for boiling surfaces exist in the literature; however, a full, physically based, and commonly accepted universal system is still missing. This paper proposes such a classification system, based on considerations of physical mechanisms underlying the nucleation process and enhancement mechanism during different stages of vapor bubble growth. It also presents an overview of recent advances in the development of enhanced boiling surfaces.

Journal of Nuclear Engineering doi: 10.3390/jne6010006

Authors: Hiroki Tanaka Takafumi Narukawa Takashi Takata

Nuclear power plant risk assessments rely on conservative deterministic criteria for core-damage determination despite significant advancements in plant response and system analyses. This study proposes a probabilistic approach to determine fuel rod fracture during loss-of-coolant accidents (LOCAs) in light-water reactors, addressing the need for more rational and realistic assessments. The methodology integrates a fuel rod fracture probability estimation model with best-estimate-plus-uncertainty analysis of plant response, utilizing the stress–strength model and Monte Carlo simulations. Both stress and strength distributions are estimated through Bayesian statistical modeling, with numerical integration techniques implemented to enhance accuracy for low-frequency events. The application of this approach to a virtual dataset demonstrated that while conventional deterministic methods indicated definitive rod fracture, our probabilistic analysis revealed a more realistic fracture probability of 15.1%. This significant finding highlights the potential reduction in assessment conservatism. The proposed methodology enables a transition from conservative binary evaluations to more realistic probabilistic assessments of core damage, providing more accurate risk insights for decision-making.

Journal of Nuclear Engineering doi: 10.3390/jne6010005

Authors: Keith E. Holbert

Marine reactors have been applied to floating nuclear power plants, naval vessels such as submarines, and civilian ships such as icebreakers. Nuclear-powered shipping is gaining increased interest because of decarbonization goals motivated by climate change. Enhanced reactor safety can potentially reduce regulatory and liability challenges to the adoption of nuclear propulsion systems for merchant ships. This gives strong impetus for reviewing past use of nuclear reactor systems in marine environments, especially from the perspective of any accident scenarios, lest planners be caught unaware of historical incidents. To that end, a loss of coolant accident (LOCA) in a Lenin icebreaker reactor in 1965 and disposal at sea of some of its damaged fuel and reactor vessel as well as the entire tri-reactor compartment is recounted.

Journal of Nuclear Engineering doi: 10.3390/jne6010004

Authors: Ugur Mertyurek Riley Cumberland William A. Wieselquist

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

Journal of Nuclear Engineering doi: 10.3390/jne6010003

Authors: Alexander Ioffe Petr Konik Konstantin Batkov

A novel concept of cold neutron source employing chessboard or staircase assemblies of high-aspect-ratio rectangular para-hydrogen moderators with well-developed and practically fully illuminated surfaces of the individual moderators is proposed. An analytic approach for calculating the brightness of para-hydrogen moderators is introduced. Because the brightness gain originates from a near-surface effect resulting from the prevailing single-collision process during thermal-to-cold neutron conversion, high-aspect-ratio rectangular cold moderators offer a significant increase, up to a factor of 10, in cold neutron brightness compared to a voluminous moderator. The obtained results are in excellent agreement with MCNP calculations. The chessboard or staircase assemblies of such moderators facilitate the generation of wide neutron beams with simultaneously higher brightness and intensity compared to a para-hydrogen-based cold neutron source made of a single moderator (either flat or voluminous) of the same cross-section. Analytic model calculations indicate that gains of up to approximately 2.5 in both brightness and intensity can be achieved compared to a source made of a single moderator of the same width. However, these gains are affected by details of the moderator–reflector assembly and should be estimated through dedicated Monte Carlo simulations, which can only be conducted for a particular neutron source and are beyond the scope of this general study. The gain reduction in our study, from a higher value to 2.5, is mostly caused by these two factors: the limited volume of the high-density thermal neutron region surrounding the reactor core or spallation target, which restricts the total length of the moderator assembly, and the finite width of moderator walls. The relatively large length of moderator assemblies results in a significant increase in pulse duration at short pulse neutron sources, making their straightforward use very problematic, though some applications are not excluded. The concept of “low-dimensionality” in moderators is explored, demonstrating that achieving a substantial increase in brightness necessitates moderators to be low-dimensional both geometrically, implying a high aspect ratio, and physically, requiring the moderator’s smallest dimension to be smaller than the characteristic scale of moderator medium (about the mean free path for thermal neutrons). This explains why additional compression of the moderator along the longest direction, effectively giving it a tube-like shape, does not result in a significant brightness increase comparable to the flattening of the moderator.

Journal of Nuclear Engineering doi: 10.3390/jne6010002

Authors: Stylianos A. Papazis

This research presents a new solution for optimizing the economics of energy produced by a hybrid power generation plant that converts nuclear, solar, and thermal energy into electricity while operating under load-following conditions. To achieve the benefits of cleaner electricity with minimal production costs, multi-criteria management decisions are applied. The investigation of a hybrid system combining nuclear, solar, and thermal energy generation demonstrates the impact of such technology on the optimal price of generated energy; the introduction of nuclear reactors in hybrid systems reduces the cost of electricity production compared to the equivalent cost of energy produced by solar systems and compared to fossil fuel thermal systems. This method can be applied to hybrid energy systems with nuclear, solar, and thermal power generation plants of various sizes and configurations, making it a useful tool for engineers, researchers, and managers in the energy sector.

Journal of Nuclear Engineering doi: 10.3390/jne6010001

Authors: Gueorgui Petkov

Risk management strives to reach the standards of theoretical systematicity and empirical precision achieved in natural science models. To this end, a set of risk-informed and performance-based standards was developed in the form of statistically validated measures. The set enables the systematic extraction by deterministic and probabilistic analysis of potentially objective risk assessments and well-defined decisions. However, much of the data and models are subjectively influenced by the uncertainty of the context in which they are related and derived. Current risk analysis contains a large amount of risk-related information, but without the context of the models, its results lack sufficient predictive and explanatory power to be a solid basis for decisions. Therefore, to model the entire site of a multi-unit nuclear power plant as an integrated system connecting facility and activity, it is necessary to consider not only the technological conditions, but also the entire site context, including human, organizational, and environmental factors. An interface tool for dynamic deterministic-probabilistic safety analysis should be used to contextualize and complement existing risk indicators, but not to replace them. This article presents the possibilities of risk contextualization for nuclear systems through the symptom-based context quantification procedure of the Performance Evaluation Teamwork method.

Journal of Nuclear Engineering doi: 10.3390/jne5040036

Authors: Dhruv Garg Ricardo Lopez Oskari Pakari Shaun D. Clarke Sara A. Pozzi

The ability to localize and image radiation sources has found use in various applications for nuclear nonproliferation practices, specifically in treaty verification, nuclear safeguards, and homeland security. Technologies that are capable of angular radiation imaging have been prevalent for years and, recently, 3D imaging technologies making use of emerging media like mixed reality have been rapidly developing and gaining popularity. Modern imaging techniques typically use a Compton camera to record coincident events and reconstruct the incident directional information of a gamma ray-emitting radiation source. However, Compton cameras are limited as they cannot obtain accurate source depth information when used for simple back projection imaging. Neutron scatter cameras are a complementary imaging technique that use double elastic scatters but also have their own limitations. This work presents a framework for multiple scatter-based particle imagers to construct 3D images and to localize a radiation source using gamma rays or fast neutrons. Specifically, localization is achieved by accounting for the position of the imagers. The imaging algorithm was validated using experimental data, measuring a 252Cf source. A three-dimensional representation of the imaging data provides a more intuitive and informative depiction of source positions and can aid in scenarios with complex environmental geometries such as when sources are in containers.

Journal of Nuclear Engineering doi: 10.3390/jne5040035

Authors: Kenneth Lee Fossum Palash Kumar Bhowmik Piyush Sabharwall

Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by evaluating experimental and computational fluid dynamics modeling approaches to support early-stage reactor system design, testing, and model evaluation. Previous studies reported in the literature for steam-flow entrainment primarily focused on gigawatt capacity pressurized water reactor (PWR) systems. However, entrainment phenomena are even more prominent for PWR-type SMRs due to their more compact integrated designs, which need further research and development. To fill the research gaps, this study provides insight by specifying the phenomena of interest by leveraging the lessons learned from past research, adopting advanced M&S techniques and advanced instrumentation and control. The findings and recommendations are applicable for evaluating steam-flow entrainment models and for designing integral effect test and separate effect test facilities for gaining reactor design approvals.

Journal of Nuclear Engineering doi: 10.3390/jne5040034

Authors: Wenjing Li Lin Xiao Lori Walters Greg Kasprick Robyn Sloan

The high-temperature low-cycle fatigue (LCF) behaviour of Incoloy 800H and its weldments with Haynes 230 and Inconel 82 filler metals, which were fabricated with the gas tungsten arc welding (GTAW) technique, was investigated and compared at 760 °C. The results revealed that the Incoloy 800H weldments showed lower fatigue lifetimes compared to the base metal. However, the weldments with the Haynes 230 filler metal demonstrated an improved fatigue life at the low strain amplitude compared to both Incoloy 800H and the weldment with the Inconel 82 filler metal. The Incoloy 800H base metal showed pronounced initial cyclic hardening with hardening factors increasing with strain amplitudes. In contrast, the weldments with Haynes 230 and Inconel 82 filler metals displayed short initial cyclic hardening and saturation stages, followed by long continuous cyclic softening. The fractography and microstructure after LCF the tests were characterized with scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Transgranular fracture with multiple crack initiations was the predominant failure mode on the fracture surfaces of both Incoloy 800 base metal and the weldments. TEM examination revealed that planar dislocation slips at the low strain amplitude evolved to wavy slips, eventually forming a cell structure at high strain amplitudes in the Incoloy 800H material as the strain amplitudes increased. However, the weld metal exhibited a planar slip mode deformation mechanism regardless of cyclic strain amplitude in the weldment specimens. The differing cyclic hardening and softening behaviours between Incoloy 800H and its weldments are attributed to the higher strength of the weldment specimens compared to the base metal. In the Incoloy 800H base material specimens, the reverse strains during LCF created wavy dislocation structures, which could not fully recover due to the non-reversible nature of the microstructure. As a result, cells or subgrains formed within the microstructure once created. In contrast, the higher strength of the weld metal in the weldment specimens significantly suppressed the formation of wavy dislocation structures, and deformation primarily manifested as planar arrays of dislocations.

Journal of Nuclear Engineering doi: 10.3390/jne5040033

Authors: Indrek Jõgi Jasper Ristkok Peeter Paris

Laser-Induced Breakdown Spectroscopy (LIBS) is one candidate for analyzing the fuel retention in ITER plasma-facing components during maintenance breaks when the reactor is filled with near atmospheric pressure nitrogen or dry air. It has been shown that using argon flow during LIBS measurements increases the LIBS signal at atmospheric pressure conditions and helps to distinguish the hydrogen isotopes. However, atmospheric pressure might be suboptimal for such LIBS measurements. The present study investigated the effect of argon or nitrogen gas at different pressures on the hydrogen Hα line emission intensity during the LIBS measurements. Laser pulses with an 8 ns width were used to ablate a small amount of a molybdenum (Mo) target with hydrogen impurity. The development of the formed plasma plume was investigated by time- and space-resolved emission spectra and photographs. Photographs showed that the plasma plume development was similar for both gases, while the total intensity of the plume was higher in argon. Space-resolved emission spectra also had stronger Hα line intensities in argon. Shorter delay times necessitated the use of lower pressures to have sufficiently narrow lines for the distinguishing of the hydrogen isotopes. At the same line widths, the line intensities were higher at lower gas pressures and in argon. Hα and Mo I line emissions were spatially separated, which suggests that the geometry of collection optics should be considered when using LIBS.

Journal of Nuclear Engineering doi: 10.3390/jne5040032

Authors: Megan N. Schiferl Jeffrey R. McLachlan Appie A. Peterson Naomi E. Marks Rebecca J. Abergel

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

Journal of Nuclear Engineering doi: 10.3390/jne5040031

Authors: Izabela Gutowska Robert Kile Brian G. Woods Nicholas R. Brown

The development of the Modular High-Temperature Gas-Cooled Reactor is a significant milestone in advanced nuclear reactor technology. One of the concerns for the reactor’s safe operation is the effects of a loss-of-flow accident (LOFA) where the coolant circulators are tripped, and forced coolant flow through the core is lost. Depending on the steam generator placement, loop or intracore natural circulation develops to help transfer heat from the core to the reactor cavity, cooling system. This paper investigates the fundamental physical phenomena associated with intracore coolant natural circulation flow in a one-sixth Computational Fluid Dynamics (CFD) model of the Oregon State University High Temperature Test Facility (OSU HTTF) following a loss-of-flow accident transient. This study employs conjugate heat transfer and steady-state flow along with an SST k-ω turbulence model to characterize the phenomenon of core channel-to-channel natural convection. Previous studies have revealed the importance of complex flow distribution in the inlet and outlet plenums with the potential to generate hot coolant jets. For this reason, complete upper and lower plenum volumes are included in the analyzed computational domain. CFD results also include parametric studies performed for a mesh sensitivity analysis, generated using the STAR-CCM+ software. The resulting channel axial velocities and flow directions support the test facility scaling analysis and similarity group distortions calculation.

Journal of Nuclear Engineering doi: 10.3390/jne5040030

Authors: Graham Medlin Ekaterina Korobkina Cole Teander Bernard Wehring Eduard Sharapov Ayman I. Hawari Paul Huffman Albert R. Young Grant Palmquist Matthew Morano Clark Hickman Thomas Rao Robert Golub

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University’s PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source’s heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources.

Journal of Nuclear Engineering doi: 10.3390/jne5040029

Authors: Adolphus Lye Jathniel Chang Sicong Xiao Keng Yeow Chung

The paper provides an introduction to the concept of Probabilistic Safety Assessment, an evaluation of its recent developments, and perspectives on the future research directions in this area. To do so, a conceptual understanding to safety assessment is first provided, followed by an introduction to what Probabilistic Safety Assessment is about. From this, the historical background and development of Probabilistic Safety Assessment in the context of nuclear safety are discussed, including a brief description and evaluation of some methods implemented to perform such analysis. After this, the paper reviews some of the recent research developments in Probabilistic Safety Assessment in the aspects of multi-unit safety assessment, dynamic Probabilistic Safety Assessment, reliability analysis, cyber-security, and policy-making. Each aspect is elaborated in detail, with perspectives provided on its potential limitations. Finally, the paper discusses research topics in six areas and challenges within the Probabilistic Safety Assessment discipline, for which further investigation might be conducted in the future. Hence, the objectives of the review paper are (1) to serve as a tutorial for readers who are new to the concept of Probabilistic Safety Assessment; (2) to provide a historical perspective on the development of the Probabilistic Safety Assessment field over the past seven decades; (3) to review the state-of-the-art developments in the use of Probabilistic Safety Assessment in the context of nuclear safety; (4) to provide an evaluative perspective on the methods implemented for Probabilistic Safety Assessment within the current literature; and (5) to provide perspectives on the future research directions that can potentially be explored, thereby also targeting the wider research community within the nuclear safety discipline towards pushing the frontiers of Probabilistic Safety Assessment research.

Journal of Nuclear Engineering doi: 10.3390/jne5040028

Authors: Borong Tang Shenfei Wang Fang Liu Fenglei Niu

Plate fuel elements, known for their compact structure and efficient cooling, are commonly used in the core of nuclear reactors. In these reactors, coolant channels are designed as rectangular narrow slits. Bubble behavior in narrow channels differs significantly from that in conventional channels. This paper investigates the vertical rise of bubbles in narrow slit channels. A gas–liquid two-phase flow experimental rig was constructed using transparent acrylic boards. A high-speed camera captured the bubble formation process during gas injection, and code implemented in Matlab was used to process the images. Numerical simulations were conducted with CFD software under identical conditions and compared with the experimental results, showing a good agreement. The results show that the experimental and simulated bubble movement velocities are in good agreement. In the experiments of this paper, when the width of the narrow gap is below 3 mm, the sidewalls exert a pronounced influence on the dynamics of bubble rise, notably altering both the velocity profile and the trajectory of the bubbles’ ascent. As the gas injection flow rate gradually increases, the bubble rising speed and trajectory change from regular to oscillatory patterns.